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Sample records for graphite fuel blocks

  1. Criteria for the selection of graphites for HTR integral block fuel elements

    International Nuclear Information System (INIS)

    Knowles, A.N.

    1980-01-01

    This paper is concerned with the special requirements for integral block fuel elements of the type first used in the Fort St. Vrain reactor. The main idea of these elements is that the carrier block and separate graphite clad fuel pins are combined into a single monolith. This combination leads to lower fabrication costs and some improvement in the thermal performance (lower temperature difference between fuel and the surface of heat transfer into the coolant). The advent of block fuel for HTRs of the Fort St. Vrain type has placed a fresh emphasis on the selection of graphite for block manufacture in respect of physical properties. This is because the temperature distributions typical of such fuelled blocks lead to shutdown stresses close to the maximum the graphite can sustain without damage. Figures presented in this paper suggest that the physical properties of the graphite can play a relatively large part in reducing such stress levels and that guidance on the key requirements for suitable specifications is therefore particularly needed by the manufacturers of fuel block graphites. While graphites for fuel blocks have this special need for combinations of physical properties which lead to low thermal and shrinkage stresses, the other characteristics must also receive attention. A low graphite cost combined with good homogeneity in the brick, so that waste minimized, are still necessary, while isotropy is also very important

  2. Calculation of thermal stresses in graphite fuel blocks

    International Nuclear Information System (INIS)

    Lejeail, Y.; Cabrillat, M.T.

    2005-01-01

    This paper presents a parametric study of temperature and thermal stress calculations inside a HTGR core graphite block, taking into account the effect of fluence on the thermal and mechanical properties, up to 4. 10 21 n/cm 2 . The Finite Element model, realized with Cast3M CEA code, includes the effects of irradiation creep, which tends to produce secondary stress relaxation. Then, the Weibull weakest link theory is recalled, evaluating the possible effects of volume, stress field distribution (loading factor), and multiaxiality for graphite-type materials, and giving the methodology to compare the stress to rupture for the structure to the one obtained from characterization, in the general case. The maximum of the Weibull stress in Finite Element calculations is compared to the value for tensile specimens. It is found that the maximum of the stress corresponds to the end of the irradiation cycle, after reactor shutdown, since both thermal conductivity and Young's modulus increase with time. However, this behaviour is partly counterbalanced by the increase of material strength with irradiation. (authors)

  3. 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)

  4. 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.

  5. 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

  6. Development of the loss coefficient correlation for cross flow between graphite fuel blocks in the core of prismatic very high temperature reactor-PMR200

    International Nuclear Information System (INIS)

    Lee, Jeong-Hun; Cho, Hyoung-Kyu; Park, Goon-Cherl

    2016-01-01

    Highlights: • Cross flow experimental data are produced with wedge-shaped and parallel gaps. • The results of a CFD analysis and experimental data are in good agreement. • Pressure loss coefficient for the cross gap between fuel blocks in PMR200 is found. • A new correlation of the cross flow loss coefficient for PMR200 is proposed. - Abstract: The core of the very high temperature reactor (VHTR) PMR200 (a prismatic modular reactor rated at 200 MW of thermal power) consists of hexagonal prismatic fuel blocks and reflector blocks made of graphite. If the core bypass flow ratio increases, the coolant channel flow is decreased and can then lower the heat removal efficiency, resulting in a locally increased fuel block temperature. The coolant channels in the fuel blocks are connected to bypass gaps by the cross gap, complicating flow distribution in the VHTR core. Therefore, reliable estimation of the bypass flow is highly important for the design and safety analysis of the VHTR core. Because of the complexity of the core geometry and gap configuration, it is challenging to predict the flow distribution in the VHTR core. To analyze this flow distribution accurately, it is necessary to determine the cross flow phenomena, and the loss coefficient across the cross gap has to be evaluated to determine the flow distribution in the VHTR core when a lumped parameter code or a flow network analysis code that uses the correlation of the loss coefficient is employed. The purpose of this paper is to develop a loss coefficient correlation applicable to the cross gap in the PMR200 core. The cross flow was evaluated experimentally using the difference between the measured inlet and outlet mass flow rates. Next, the applicability of a commercial computational fluid dynamics (CFD) code, CFX 15, was confirmed by comparing the experimental data and CFD analysis results. To understand the cross flow phenomena, the loss coefficient was evaluated; in the high Reynolds number region

  7. Development of the loss coefficient correlation for cross flow between graphite fuel blocks in the core of prismatic very high temperature reactor-PMR200

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong-Hun, E-mail: huny12@snu.ac.kr; Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr; Park, Goon-Cherl, E-mail: parkgc@snu.ac.kr

    2016-10-15

    Highlights: • Cross flow experimental data are produced with wedge-shaped and parallel gaps. • The results of a CFD analysis and experimental data are in good agreement. • Pressure loss coefficient for the cross gap between fuel blocks in PMR200 is found. • A new correlation of the cross flow loss coefficient for PMR200 is proposed. - Abstract: The core of the very high temperature reactor (VHTR) PMR200 (a prismatic modular reactor rated at 200 MW of thermal power) consists of hexagonal prismatic fuel blocks and reflector blocks made of graphite. If the core bypass flow ratio increases, the coolant channel flow is decreased and can then lower the heat removal efficiency, resulting in a locally increased fuel block temperature. The coolant channels in the fuel blocks are connected to bypass gaps by the cross gap, complicating flow distribution in the VHTR core. Therefore, reliable estimation of the bypass flow is highly important for the design and safety analysis of the VHTR core. Because of the complexity of the core geometry and gap configuration, it is challenging to predict the flow distribution in the VHTR core. To analyze this flow distribution accurately, it is necessary to determine the cross flow phenomena, and the loss coefficient across the cross gap has to be evaluated to determine the flow distribution in the VHTR core when a lumped parameter code or a flow network analysis code that uses the correlation of the loss coefficient is employed. The purpose of this paper is to develop a loss coefficient correlation applicable to the cross gap in the PMR200 core. The cross flow was evaluated experimentally using the difference between the measured inlet and outlet mass flow rates. Next, the applicability of a commercial computational fluid dynamics (CFD) code, CFX 15, was confirmed by comparing the experimental data and CFD analysis results. To understand the cross flow phenomena, the loss coefficient was evaluated; in the high Reynolds number region

  8. Monte Carlo calculation of standard graphite block

    International Nuclear Information System (INIS)

    Ljubenov, V.

    2000-01-01

    This paper presents results of calculation of neutron flux space and energy distribution in the standard graphite block (SGB) obtained by the MCNP TM code. VMCCS nuclear data library, based on the ENDF / B-VI release 4 evaluation file, is used. MCNP model of the SGB considers detailed material, geometric and spectral properties of the neutron source, source carrier, graphite moderator medium, aluminium foil holders and proximate surrounding of SGB Geometric model is organised to provide the simplest homogeneous volume cells in order to obtain the maximum acceleration of neutron history tracking (author)

  9. Integral-fuel blocks

    International Nuclear Information System (INIS)

    Cunningham, C.; Simpkin, S.D.

    1975-01-01

    A prismatic moderator block is described which has fuel-containing channels and coolant channels disposed parallel to each other and to edge faces of the block. The coolant channels are arranged in rows on an equilateral triangular lattice pattern and the fuel-containing channels are disposed in a regular lattice pattern with one fuel-containing channel between and equidistant from each of the coolant channels in each group of three mutually adjacent coolant channels. The edge faces of the block are parallel to the rows of coolant channels and the channels nearest to each edge face are disposed in two rows parallel thereto, with one of the rows containing only coolant channels and the other row containing only fuel-containing channels. (Official Gazette)

  10. Graphite behaviour in relation to the fuel element design

    Energy Technology Data Exchange (ETDEWEB)

    Everett, M. R. [OECD High Temperature Reactor Project Dragon, Winfrith (United Kingdom); Manzel, R. [OECD High Temperature Reactor Project Dragon, Winfrith (United Kingdom); Blackstone, R. [Reactor Centrum, Petten (Netherlands); Delle, W. [Kernforschungsanlage, Juelich (Germany); Lungagnani, V. [Joint Nuclear Research Centre, Euratom, Petten (Netherlands); Krefeld, R. [Joint Nuclear Research Centre, Euratom, Petten (Netherlands)

    1969-09-01

    The first designs of H.T.R. power reactors will probably use a Gilsocarbon based graphite for both the moderator/carrier blocks and for the fuel tubes. The initial physical properties and changes of dimensions, thermal expansion coefficient, Young*s modulus, and thermal conductivity on irradiation of Gilsocarbon graphites to typical reactor dwell-time fast neutron doses of 4 * 1021 cm -2 Ni dose Dido equivalent are given and values for the irradiation creep constant are presented. The influence of these property changes and those of chemical corrosion are considered briefly in relation to the present fuel element designs. The selection of an eventual less costly replacement graphite for Gilsocarbon graphite is discussed in terms of materials properties.

  11. Block fuel element for gas-cooled high temperature reactors

    International Nuclear Information System (INIS)

    Hrovat, M.F.

    1978-01-01

    The invention concerns a block fuel element consisting of only one carbon matrix which is almost isotropic of high crystallinity into which the coated particles are incorporated by a pressing process. This block element is produced under isostatic pressure from graphite matrix powder and coated particles in a rubber die and is subsequently subjected to heat treatment. The main component of the graphite matrix powder consists of natural graphite powder to which artificial graphite powder and a small amount of a phenol resin binding agent are added

  12. 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

  13. Development and engineering plan for graphite spent fuels conditioning program

    International Nuclear Information System (INIS)

    Bendixsen, C.L.; Fillmore, D.L.; Kirkham, R.J.; Lord, D.L.; Phillips, M.B.; Pinto, A.P.; Staiger, M.D.

    1993-09-01

    Irradiated (or spent) graphite fuel stored at the Idaho Chemical Processing Plant (ICPP) includes Fort St. Vrain (FSV) reactor and Peach Bottom reactor spent fuels. Conditioning and disposal of spent graphite fuels presently includes three broad alternatives: (1) direct disposal with minimum fuel packaging or conditioning, (2) mechanical disassembly of spent fuel into high-level waste and low-level waste portions to minimize geologic repository requirements, and (3) waste-volume reduction via burning of bulk graphite and other spent fuel chemical processing of the spent fuel. A multi-year program for the engineering development and demonstration of conditioning processes is described. Program costs, schedules, and facility requirements are estimated

  14. 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.)

  15. Analytical study of stress and deformation of HTR fuel blocks

    International Nuclear Information System (INIS)

    Tanaka, M.

    1982-01-01

    A two-dimensional finite element computer code named HANS-GR has been developed to predict the mechanical behavior of the graphite fuel blocks with realistic material properties and core environment. When graphite material is exposed to high temperature and fast neutron flux of high density, strains arise due to thermal expansion, irradiation-induced shrinkage and creep. Thus stresses and distortions are induced in the fuel block in which there are spatial variation of these strains. The analytical method used in the program to predcit these induced stresses and distortions by finite element method is discussed. In order to illustrate the versatility of the computer code, numerical results of two example analyses of the multi-hole type fuel elements in the VHTR Reactor are given. Two example analyses presented are those concerning the stresses in fuel blocks with control rod holes and distortions of the fuel blocks at the periphery of the reactor core. It is considered these phenomena should be carefully examined when the multi-hole type fuel elements are applied to VHTR. It is assured that the predicted mechanical behavior of the graphite components is strongly dependent on the material properties used and obtaining the reliable material property is important to make the analytical prediction a reliable one

  16. Verification of thermal-irradiation stress analytical code VIENUS of graphite block

    International Nuclear Information System (INIS)

    Iyoku, Tatsuo; Ishihara, Masahiro; Shiozawa, Shusaku; Shirai, Hiroshi; Minato, Kazuo.

    1992-02-01

    The core graphite components of the High Temperature Engineering Test Reactor (HTTR) show both the dimensional change (irradiation shrinkage) and creep behavior due to fast neutron irradiation under the temperature and the fast neutron irradiation conditions of the HTTR. Therefore, thermal/irradiation stress analytical code, VIENUS, which treats these graphite irradiation behavior, is to be employed in order to design the core components such as fuel block etc. of the HTTR. The VIENUS is a two dimensional finite element viscoelastic stress analytical code to take account of changes in mechanical properties, thermal strain, irradiation-induced dimensional change and creep in the fast neutron irradiation environment. Verification analyses were carried out in order to prove the validity of this code based on the irradiation tests of the 8th OGL-1 fuel assembly and the fuel element of the Peach Bottom reactor. This report describes the outline of the VIENUS code and its verification analyses. (author)

  17. 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

  18. 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.

  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. 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

  1. Application of 3D coupled code ATHLET-QUABOX/CUBBOX for RBMK-1000 transients after graphite block modernization

    Energy Technology Data Exchange (ETDEWEB)

    Samokhin, Aleksei [Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS), Moscow (Russian Federation); Zilly, Matias [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-11-15

    This work describes the application and the results of transient calculations for the RBMK-1000 with the coupled code system ATHLET 2.2A-QUABOX/CUBBOX which was developed in GRS. Within these studies the planned modernization of the graphite blocks of the RBMK-1000 reactor is taken into account. During the long-term operation of the uranium-graphite reactors RBMK-1000, a change of physical and mechanical properties of the reactor graphite blocks is observed due to the impact of radiation and temperature effects. These have led to a deformation of the reactor graphite columns and, as a result, a deformation of the control and protection system (CPS) and of fuel channels. Potentially, this deformation can lead to problems affecting the smooth movement of the control rods in the CPS channels and problems during the loading and unloading of fuel assemblies. The present paper analyzes two reactivity insertion transients, each taking into account three graphite removal scenarios. The presented work is directly connected with the modernization program of the RBMK- 1000 reactors and has an important contribution to the assessment of the safety-relevant parameters after the modification of the core graphite blocks.

  2. (Fuel, fission product, and graphite technology)

    Energy Technology Data Exchange (ETDEWEB)

    Stansfield, O.M.

    1990-07-25

    Travel to the Forschungszentrum (KFA) -- Juelich described in this report was for the purpose of participating in the annual meeting of subprogram managers for the US/DOE Umbrella Agreement for Fuel, Fission Product, and Graphite Technology. At this meeting the highlights of the cooperative exchange were reviewed for the time period June 1989 through June 1990. The program continues to contribute technology in an effective way for both countries. Revision 15 of the Subprogram Plan will be issued as a result of the meeting. There was interest expressed by KFA management in the level of support received from the NPR program and in potential participation in the COMEDIE loop experiment being conducted at the CEA.

  3. Low temperature chemical processing of graphite-clad nuclear fuels

    Science.gov (United States)

    Pierce, Robert A.

    2017-10-17

    A reduced-temperature method for treatment of a fuel element is described. The method includes molten salt treatment of a fuel element with a nitrate salt. The nitrate salt can oxidize the outer graphite matrix of a fuel element. The method can also include reduced temperature degradation of the carbide layer of a fuel element and low temperature solubilization of the fuel in a kernel of a fuel element.

  4. Effect of the Heat Treatment on the Graphite Matrix of Fuel Element for HTGR

    International Nuclear Information System (INIS)

    Lee, Chungyong; Lee, Seungjae; Suh, Jungmin; Jo, Youngho; Lee, Youngwoo; Cho, Moonsung

    2013-01-01

    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 for 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 Phenol as a binder were chosen and mixed with each other, formed and heated for the compressive strength test. The objective of this research is to optimize the kinds and composition of the mixed graphite and the forming process by evaluating the compressive strength before/after heat treatment (carbonization of binder). In this study, the effect of heat treatment on graphite matrix was studied in terms of the density and the compressive strength. The size (diameter and length) of pellet is increased by heat treatment. Due to additional weight reduction and swelling (length and diameter) of samples the density of graphite pellet is decreased from about 2.0 to about 1.7g/cm 3 . From the mechanical test results, the compressive strength of graphite pellets was related to the various conditions such as the contents of binder, the kinds of graphite and the heat treatment. Both the green pellet and the heat treated pellet, the compressive strength of G+S+P pellets is relatively higher than that of R+S+P pellets. To optimize fuel element matrix, the effect of Phenol and other binders, graphite composition and the heat treatment on the mechanical properties will be deeply investigated for further study

  5. Ageing Management of Beryllium and Graphite Blocks in Research Reactor MARIA

    Energy Technology Data Exchange (ETDEWEB)

    Golab, A. [National Centre for Nuclear Research, Warsaw (Poland)

    2013-07-01

    In the paper the phenomenon of beryllium moderator poisoning by thermal neutron absorption and the method and results of this phenomenon control is presented. Also the phenomenon of graphite blocks damage due to fast neutrons accumulation and the methods and results of this process supervising is described. These methods refer especially to: visual inspection of their state and radiography of graphite blocks. Special attention is paid to permanent estimate of fast neutron fluency accumulated in blocks and methods of their shuffling in the reactor core. The shuffling makes possible to increase the lifetime of beryllium and graphite blocks and decrease the cost of reactor operation.

  6. Corrosion of graphite composites in phosphoric acid fuel cells

    Science.gov (United States)

    Christner, L. G.; Dhar, H. P.; Farooque, M.; Kush, A. K.

    1986-01-01

    Polymers, polymer-graphite composites and different carbon materials are being considered for many of the fuel cell stack components. Exposure to concentrated phosphoric acid in the fuel cell environment and to high anodic potential results in corrosion. Relative corrosion rates of these materials, failure modes, plausible mechanisms of corrosion and methods for improvement of these materials are investigated.

  7. Thermal stress analysis of HTGR fuel and control rod fuel blocks in the HTGR in-block carbonization and annealing furnace

    International Nuclear Information System (INIS)

    Gwaltney, R.C.; McAfee, W.J.

    1977-01-01

    A new approach that utilizes the equivalent solid plate method has been applied to the thermal stress analysis of HTGR fuel and control rod fuel blocks. Cases were considered where these blocks, loaded with reprocessed HTGR fuel pellets, were being cured at temperatures up to 1800 0 C. A two-dimensional segment of a fuel block cross section including fuel, coolant holes, and graphite matrix was analyzed using the ORNL HEATING3 heat transfer code to determine the temperature-dependent effective thermal conductivity for the perforated region of the block. Using this equivalent conductivity to calculate the temperature distributions through different cross sections of the blocks, two-dimensional thermal-stress analyses were performed through application of the equivalent solid plate method. In this approach, the perforated material is replaced by solid homogeneous material of the same external dimensions but whose material properties have been modified to account for the perforations

  8. Thermal-stress analysis of HTGR fuel and control rod fuel blocks in in-block carbonization and annealing furnace

    International Nuclear Information System (INIS)

    Gwaltney, R.C.; McAfee, W.J.

    1977-01-01

    A new method for performing thermal stress analyses in structures with multiple penetrations was applied to these analyses. This method couples the development of an equivalent thermal conductivity for the blocks, a technique that has been used extensively for modeling the thermal characteristics of reactor cores, with the use of the equivalent solid plate method for stress analysis. Using this equivalent thermal conductivity, which models as one material the heat transfer characteristics of the fuel, coolant, and graphite two-dimensional, steady-state thermal analyses of the fuel and control rod fuel blocks were performed to establish all temperature boundaries required for the stress analyses. In applying the equivalent solid plate method, the region of penetrations being modeled was replaced by a pseudo material having the same dimensions but whose materials properties were adjusted to account for the penetration. The peak stresses and strains were determined by applying stress and strain intensification factors to the calculated distributions. The condition studied was where the blocks were located near the center of the furnace. In this position, the axial surface of the block is heated near one end and cooled near the other. The approximate axial surface temperatures ranged from 1521 0 C at both the heated and the cooled ends to a peak of 1800 0 C near the center. Five specific cases were analyzed: plane (two-dimensional thermal, plane stress strain) analyses of each end of a standard fuel block (2 cases), plane analyses of each end of a control rod fuel block (2 cases), and a two-dimensional analysis of a fuel block treated as an axisymmetric cylind

  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. Computation of deformations and stresses in graphite blocks for HTR core survey purposes

    International Nuclear Information System (INIS)

    Besdo, Dieter; Theymann, W.

    1975-01-01

    Stresses and deformations in graphite fuel elements for HTRs are caused by the temperature distribution and by irradiation under influence of creep, shrinking, thermal strains, and elastic deformations. The global deformations and the stress distribution in a prismatic fuel-element containing regularly distributed axial holes for the coolant flow and the fuel sticks, can be computed in the following manner: the block with its holes is treated as an effective homogeneous continuum with an equivalent global behaviour. Assuming that the fourth-order-tensor of the elastic constants is proportional to the corresponding tensor in the constitutive equations for creep, only the effective strains are of interest. The values of temperature and dose may be given in n points of the block at certain points of time. Then, the inelastic nonthermal strains are integrated by a Runge-Kutta-procedure in the n points. When interpolated and combined with thermal strains, they are incompatible. Hence, they produce elastic deformations which cause creep and can be computed by use of a Ritz-polynomial-series by help of a specific principle of the minimum of potential energy. Excessive computing time can be avoided easily since the influence of the local variation of the elastic constants within the block is almost negligible and, therefore, of practically no importance for the determination of the elastic strains. By this reason some matrices can be calculated a priori, and the elastic deformations are obtained by multiplications of these matrices rather than inversions. Therefore, this method is particularly suited for the computation of deformations and stresses for reactor core survey purposes where a large number (up to 7000 blocks) have to be treated

  11. Characterization of graphite-matrix pulsed reactor fuels

    International Nuclear Information System (INIS)

    Karnes, C.H.; Marion, R.H.

    1976-01-01

    The performance of the Annular Core Pulsed Reactor (ACPR) is being upgraded in order to accommodate higher fluence experiments for fast reactor fuel element transient and safety studies. The increased fluence requires a two-zone core with the inner zone containing fuel having a high enthalpy and the capability of withstanding very high temperatures during both pulsed and steady state operation. Because the fuel is subjected to a temperature risetime of 2 to 5 ms and to a large temperature difference across the diameter, fracture due to thermal stresses is the primary failure mode. One of the fuels considered for the high enthalpy inner region is a graphite-matrix fuel containing a dispersion of uranium--zirconium carbide solid solution particles. A program was initiated to optimize the development of this class of fuel. This summary presents results on formulations of fuel which have been fabricated by the Materials Technology Group of the Los Alamos Scientific Laboratory

  12. 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)

  13. Effect of eccentric location of the RBMK CPS displacer graphite block in the shielding sheath

    International Nuclear Information System (INIS)

    Dostov, A.I.

    2001-01-01

    Temperature conditions and accumulation of Wigner energy in the graphite block of the RBMK reactor CPS (control power system) displacer is examined. It is shown, that at eccentric location of the block in the shielding sheath average temperature of the block drops sharply. Due to the design demerit quantity of the stored energy in the block may be so great, that its release will result in melting of the displacer tube. (author)

  14. Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Luther, Erik; Rooyen, Isabella van; Leckie, Rafael; Papin, Pallas; Nelson, Andrew; Hunter, James

    2015-03-01

    In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabrication must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.

  15. Improving methodology in open vessel digestion with a graphite heating block (T7)

    International Nuclear Information System (INIS)

    Kainrath, P.; Conrads, B.; Ross, A.

    2002-01-01

    Full text: Open block digestion systems have been very popular in environmental analysis over the past decades, but have consistently suffered from the major drawback of their sensitivity against corrosion and the subsequent risk of contamination. Therefore block digestion systems have not been considered state-of-the-art technology in trace and ultra trace sample preparation. Graphite block digestion systems are well established in North America and are recently becoming more frequently considered in Europe. These systems overcome the deficiencies of the traditional systems, made from stainless steel or aluminum, because the block is manufactured from graphite and typically coated with a fluoro-polymer to present the possibility metallic contamination from the surface of the system during the handling of the samples. Graphite block systems present an alternative to the current mainstream technology of open and closed vessel microwave assisted digestion systems, as they allow large numbers of samples to be digested simultaneously, thus overcoming one of the major weaknesses of closed vessel systems. More recently a number of improvements in the technology has been developed for graphite block digestion systems and studies have been performed to evaluate the effects of such improvements. The paper presented will deal with the technological improvements: monitoring and control of sample temperature vs. monitoring of block temperature, elimination of cross contamination effects during open vessel block digestion, evaporation of samples for pre-concentration or multiple digestion steps, addressing the needs of various labs and applications for block digesters. The effects of those developments will be discussed; application examples and finally an outlook into possible future trends for graphite block digestion systems will be given. (author)

  16. Improved graphite matrix for coated-particle fuel

    International Nuclear Information System (INIS)

    Schell, D.H.; Davidson, K.V.

    1978-10-01

    An experimental process was developed to incorporate coated fuel particles in an extruded graphite matrix. This structure, containing 41 vol% particles, had a high matrix density, >1.6 g/cm 3 , and a matrix conductivity three to four times that of a pitch-injected fuel rod at 1775 K. Experiments were conducted to determine the uniformity of particle loadings in extrusions. Irradiation specimens were supplied for five tests in the High-Fluence Isotope Reactor at the Oak Ridge National Laboratory

  17. Irradiation-induced dimensional changes of fuel compacts and graphite sleeves of OGL-1 fuel assemblies

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Minato, Kazuo; Kobayashi, Fumiaki; Tobita, Tsutomu; Kikuchi, Teruo; Kurobane, Shiro; Adachi, Mamoru; Fukuda, Kousaku

    1988-06-01

    Experimental data are summarized on irradiation-induced dimensional changes of fuel compacts and graphite sleeves of the first to ninth OGL-1 fuel assemblies. The range of fast-neutron fluence is up to 4 x 10 24 n/m 2 (E > 0.18 MeV); and that of irradiation temperature is 900 - 1400 deg C for fuel compacts and 800 - 1050 deg C for graphite sleeves. The dimensional change of the fuel compacts was shrinkage under these test conditions, and the shrinkage fraction increased almost linearly with fast-neutron fluence. The shrinkage fraction of the fuel compacts was larger by 20 % in the axial direction than in the radial direction. Influence of the irradiation temperature on the dimensional-change behavior of the fuel compacts was not observed clearly; presumably the influence was hidden by scatter of the data because of low level of the fast-neutron fluence and the resultant small dimensional changes. (author)

  18. Surface-reconstructed graphite nanofibers as a support for cathode catalysts of fuel cells.

    Science.gov (United States)

    Gan, Lin; Du, Hongda; Li, Baohua; Kang, Feiyu

    2011-04-07

    Graphite nanofibers (GNFs), on which surface graphite edges were reconstructed into nano-loops, were explored as a cathode catalyst support for fuel cells. The high degree of graphitization, as well as the surface-reconstructed nano-loops that possess topological defects for uniform metal deposition, resulted in an improved performance of the GNF-supported Pt catalyst.

  19. 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.

  20. Process for the production of prismatic graphite molded articles for high temperature fuel elements

    International Nuclear Information System (INIS)

    Huschka, H.; Rachor, L.; Hrovat, M.; Wolff, W.

    1976-01-01

    Prismatic graphite molded objects for high temperature fuel elements are prepared by producing the outer geometry and the holes for cooling channels and for receiving fuel and fertile materials in the formation of the carbon object

  1. 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

  2. FSV experience in support of the GT-MHR reactor physics, fuel performance, and graphite

    International Nuclear Information System (INIS)

    Baxter, A.M.; McEachern, D.; Hanson, D.L.; Vollman, R.E.

    1994-11-01

    The Fort St. Vrain (FSV) power plant was the most recent operating graphite-moderated, helium-cooled nuclear power plant in the United States. Many similarities exist between the FSV design and the current design of the GT-MHR. Both designs use graphite as the basic building blocks of the core, as structural material, in the reflectors, and as a neutron moderator. Both designs use hexagonal fuel elements containing cylindrical fuel rods with coated fuel particles. Helium is the coolant and the power densities vary by less than 5%. Since material and geometric properties of the GT-MHR core am very similar to the FSV core, it is logical to draw upon the FSV experience in support of the GT-MHR design. In the Physics area, testing at FSV during the first three cycles of operation has confirmed that the calculational models used for the core design were very successful in predicting the core nuclear performance from initial cold criticality through power operation and refueling. There was excellent agreement between predicted and measured initial core criticality and control rod positions during startup. Measured axial flux distributions were within 5% of the predicted value at the peak. The isothermal temperature coefficient at zero power was in agreement within 3%, and even the calculated temperature defect over the whole operating range for cycle 3 was within 8% of the measured defect. In the Fuel Performance area, fuel particle coating performance, and fission gas release predictions and an overall plateout analysis were performed for decommissioning purposes. A comparison between predicted and measured fission gas release histories of Kr-85m and Xe-138 and a similar comparison with specific circulator plateout data indicated good agreement between prediction and measured data. Only I-131 plateout data was overpredicted, while Cs-137 data was underpredicted

  3. Fuel elements for high temperature reactors having special suitability for reuse of the structural graphite

    International Nuclear Information System (INIS)

    Huschka, H.; Herrmann, F.J.

    1976-01-01

    There are prepared fuel elements for high temperature reactors from which the fuel zone can be removed from the structural graphite after the burnup of the fissile material has taken place so that the fuel element can be filled with new fuel and again placed in the reactor by having the strength of the matrix in the fuel zone sufficient for binding the embedded coated fuel particles but substantially less than the strength of the structural graphite whereby by the action of force it can be easily split up without destroying the particles

  4. Studies on design principles and criteria of fuels and graphites for experimental multi-purpose very high temperature reactor

    International Nuclear Information System (INIS)

    Arai, Taketoshi; Sato, Sadao; Tani, Yutaro

    1977-12-01

    Design principles and criteria of fuels and graphites have been studied to determine the main design parameters of a reference core MARK-III of the Experimental Multi-purpose Very High Temperature Reactor. The present status of research and development for HTGR fuels and graphites is reviewed from a standpoint of their integrity and safety aspects, and is compared to the specific design requirements for the VHTR fuels and graphites. Consequently, reasonable materials specifications, safety criteria and design analysis methods are presented for coated fuel particle, fuel compact, graphite sleeve, core support graphite and neutron absorber material. These design principles and criteria will be refined by further experimental investigations. (auth.)

  5. Fort St. Vrain graphite site mechanical separation concept selection

    International Nuclear Information System (INIS)

    Berry, S.M.

    1993-09-01

    One of the alternatives to the disposal of the Fort St. Vrain (FSV) reactor spent nuclear fuel involves the separation of the fuel rods composed of compacts from the graphite fuel block assembly. After the separation of these two components, the empty graphite fuel blocks would be disposed of as a low level waste (provided the appropriate requirements are met) and the fuel compacts would be treated as high level waste material. This report deals with the mechanical separation aspects concerning physical disassembly of the FSV graphite fuel element into the empty graphite fuel blocks and fuel compacts. This report recommends that a drilling technique is the preferred choice for accessing the, fuel channel holes and that each hole is drilled separately. This report does not cover any techniques or methods to separate the triso fuel particles from the graphite matrix of the fuel compacts

  6. An Experiment on the Carbonization of Fuel Compact Matrix Graphite for HTGR

    International Nuclear Information System (INIS)

    Lee, Young Woo; Kim, Joo Hyoung; Cho, Moon Sung

    2012-01-01

    The fuel element for HTGR is manufactured by mixing coated fuel particles with matrix graphite powder and forming into either pebble type or cylindrical type compacts depending on their use in different HTGR cores. The coated fuel particle, the so-called TRISO particle, consists of 500-μm spherical UO 2 particles coated with the low density buffer Pyrolytic Carbon (PyC) layer, the inner and outer high density PyC layer and SiC layer sandwiched between the two inner and outer PyC layers. The coated TRISO particles are mixed with a properly prepared matrix graphite powder, pressed into a spherical shape or a cylindrical compact, and finally heat-treated at about 1800 .deg. C. These fuel elements can have different sizes and forms of compact. The basic steps for manufacturing a fuel element include preparation of graphite matrix powder, over coating the fuel particles, mixing the fuel particles with a matrix powder, carbonizing green compact, and the final high-temperature heat treatment of the carbonized fuel compact. The carbonization is a process step where the binder that is incorporated during the matrix graphite powder preparation step is evaporated and the residue of the binder is carbonized during the heat treatment at about 1073 K, In order to develop a fuel compact fabrication technology, and for fuel matrix graphite to meet the required material properties, it is of extreme importance to investigate the relationship among the process parameters of the matrix graphite powder preparation, fabrication parameters of fuel element green compact and the carbonization condition, which has a strong influence on further steps and the material properties of fuel element. In this work, the carbonization behavior of green compact samples prepared from the matrix graphite powder mixtures with different binder materials was investigated in order to elucidate the behavior of binders during the carbonization heat treatment by analyzing the change in weight, density and its

  7. 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

  8. Distribution of 60Co and 54Mn in graphite material of irradiated HTGR fuel assemblies

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Kikuchi, Teruo; Kobayashi, Fumiaki; Minato, Kazuo; Fukuda, Kousaku; Ikawa, Katsuichi; Iwamoto, Kazumi

    1984-05-01

    Distribution of 60 Co and 54 Mn was measured in the graphite sleeves and blocks of the third and fourth HTGR fuel assemblies irradiated in the Oarai Gas Loop-1 (OGL-1), which is a high temperature inpile gas loop installed in the Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Research Institute (JAERI). Axial and circumferential profiles were obtained by gamma spectrometry, and radial profiles by lathe sectioning with gamma spectrometry. Distribution of 60 Co is in good agreement with that of thermal neutron flux, and the Co content in the graphite is estimated to be -- 1 x 10 -9 in weight fraction. Concentration of 54 Mn decreases toward the axial center in its axial profile, and radially is almost uniform inside and appreciably higher at free surfaces. An estimated Fe content of --10 -8 in wight fraction is smaller by two orders of magnitude than that from chemical analysis. Higher concentraion of 60 Co and 54 Mn at the free surfaces suggests the importance of transportation process of these nuclides in the coolant loop. (author)

  9. Graphite for high-temperature reactors

    International Nuclear Information System (INIS)

    Hammer, W.; Leushacke, D.F.; Nickel, H.; Theymann, W.

    1976-01-01

    The different graphites necessary for HTRs are being developed, produced and tested within the Federal German ''Development Programme Nuclear Graphite''. Up to now, batches of the following graphite grades have been manufactured and fully characterized by the SIGRI Company to demonstrate reproducibility: pitch coke graphite AS2-500 for the hexagonal fuel elements and exchangeable reflector blocks; special pitch coke graphite ASI2-500 for reflector blocks of the pebble-bed reactor and as back-up material for the hexagonal fuel elements; graphite for core support columns. The material data obtained fulfill most of the requirements under present specifications. Production of large-size blocks for the permanent side reflector and the core support blocks is under way. The test programme covers all areas important for characterizing and judging HTR-graphites. In-pile testing comprises evaluation of the material for irradiation-induced changes of dimensions, mechanical and thermal properties - including behaviour under temperature cycling and creep behaviour - as well as irradiating fuel element segments and blocks. Testing out-of-pile includes: evaluation of corrosion rates and influence of corrosion on strength; strength measurements; including failure criteria. The test programme has been carried out extensively on the AS2-graphite, and the results obtained show that this graphite is suitable as HTGR fuel element graphite. (author)

  10. A Sub-channel Analysis of a VHTR Fuel Block with Tin Gap-Filler

    International Nuclear Information System (INIS)

    Cho, Chung Ho; Kim, Yong Hee; Yi, Yong Sun; Kim, Hong Pyo

    2005-01-01

    In the Nuclear Hydrogen Development and Demonstration (NHDD) project, two types of VHTRs (Very High Temperature Reactors), prismatic or pebble bed, are under investigation as the nuclear heat source for hydrogen production. In general, the targeted coolant outlet temperature of VHTR is 950∼1000 .deg. C and the maximum allowable fuel temperature is 1250 .deg. C during the normal operation. In the case of the prismatic reactor (PMR), conventional fuel designs result in a small margin in the maximum fuel temperature. This is one of the biggest demerits of the prismatic type In this paper, a technique of lowering the maximum fuel temperature is suggested. The PMR fuel assembly is comprised of many coolant holes and fuel channels. Cylindrical fuel compacts are stacked inside the fuel channel. Consequently, there should be a physical gap between the fuel compact and graphite block, which is filled with the He gas in the conventional design. The heat transfer coefficient of the He gap is very poor, and this increases the fuel temperature substantially. In the proposed design measure, the gap is filled with a liquid metal, tin (Sn) that has a very high thermal conductivity. The effects of tin in the gap with gap distance variation in the viewpoint of thermal hydraulics are quantitatively discussed. Also, the effects of the variations of the axial power distribution are discussed

  11. Thermal-stress analysis of HTGR fuel and control rod fuel blocks in in-block carbonization and annealing furnace

    International Nuclear Information System (INIS)

    Gwaltney, R.C.; McAfee, W.J.

    1977-01-01

    The equivalent solid plate method, in conjunction with two-dimensional plane stress and plane strain analyses, was used in assessing the thermal stress behavior of HTGR fuel and control rod fuel blocks. For the control rod fuel blocks, particular attention was given to ascertaining the effects of the reserve shutdown hole and the control rod channel holes. The assumed safety factor of 2 on the failure criteria was considered adequate to account for neglecting the axial temperature gradient in the plane analyses of the ends of the blocks. The analyses indicated that the maximum calculated tensile stress values were smaller than the criteria values except for the plane strain analysis of the control rod fuel block end surfaces and the axisymmetric analysis of the fuel block as a circular cylinder. However, most of the maximum calculated strain values were greater than the criteria values

  12. Thermal analysis of W VII-AS limiter system and presentation of a graphite-block concept

    International Nuclear Information System (INIS)

    Mukherjee, S.; Grigull, P.

    1989-01-01

    A 2D-finite element thermal analysis of the initial W VII-AS limiter system has been performed and is discussed. Furhter to this analysis a graphite block concept is presented. This concept has been numerically analyzed for applications as a limiter in plasma and nuclear fusion experimental devices. The results are described in this paper. This block concept seems to be also applicable to first wall and divertor designs; the graphite elements could be replaced by ceramic ones. (author). 10 refs.; 13 figs

  13. HTGR Fuels and Core Development Program. Quarterly progress report for the period ending August 31, 1977. [Graphite and fuel irradiation; fission product release

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    The work reported includes studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and data are presented.

  14. 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

  15. Comparison of Material Behavior of Matrix Graphite for HTGR Fuel Elements upon Irradiation: A literature Survey

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The fuel elements for the HTGRs (i.e., spherical fuel element in pebble-bed type core design and fuel compact in prismatic core design) consists of coated fuel particles dispersed and bonded in a closely packed array within a carbonaceous matrix. This matrix is generally made by mixing fully graphitized natural and needle- or pitchcoke originated powders admixed with a binder material (pitch or phenolic resin), The resulting resinated graphite powder mixture, when compacted, may influence a number of material properties as well as its behavior under neutron irradiation during reactor operation. In the fabrication routes of these two different fuel element forms, different consolidation methods are employed; a quasi-isostatic pressing method is generally adopted to make pebbles while fuel compacts are fabricated by uni-axial pressing mode. The result showed that the hardness values obtained from the two directions showed an anisotropic behavior: The values obtained from the perpendicular section showed much higher micro hardness (176.6±10.5MPa in average) than from the parallel section ((125.6±MPa in average). This anisotropic behavior was concluded to be related to the microstructure of the matrix graphite. This may imply that the uni-axial pressing method to make compacts influence the microstructure of the matrix and hence the material properties of the matrix graphite.

  16. Influence of graphite discs, chamfers, and plenums on temperature distributions in high burnup fuel

    International Nuclear Information System (INIS)

    Ranger, A.; Tayal, M.; Singh, P.

    1990-04-01

    Previous studies have demonstrated the desirability to increase the fuel burnups in CANDU reactors from 7-9 GW.d/t to 21 GW.d/t. At high burnups, one consideration in fuel integrity is fission gas pressure, which is predicted to reach about 13 MPa. The gas pressure can be kept below the coolant pressure (about 10 MPa) via a variety of options such as bigger chamfers, deeper dishes, central hole, and plenums. However, it is important to address the temperature perturbations produced by the bigger chamfers and plenums which in turn, affect the gas pressure. Another consideration in fuel integrity is to reduce the likelihood of fuel failures via environmentally assisted cracking. Insertion of graphite discs between neighbouring pellets will lower the pellet temperatures, hence, lower fission gas release and lower expansion of the pellet. Therefore, it is desired to quantify the effect of graphite discs on pellet temperatures. Thermal analyses of different fuel element geometries: with and without chamfers, graphite discs, and plenums were performed. The results indicate that the two-dimensional distributions of temperatures due to the presence of chamfers, graphite discs, or plenums can have a significant impact on the integrity of high burnup fuel as we have been able to quantify in this paper

  17. Method of producing exfoliated graphite composite compositions for fuel cell flow field plates

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-04-08

    A method of producing an electrically conductive composite composition, which is particularly useful for fuel cell bipolar plate applications. The method comprises: (a) providing a supply of expandable graphite powder; (b) providing a supply of a non-expandable powder component comprising a binder or matrix material; (c) blending the expandable graphite with the non-expandable powder component to form a powder mixture wherein the non-expandable powder component is in the amount of between 3% and 60% by weight based on the total weight of the powder mixture; (d) exposing the powder mixture to a temperature sufficient for exfoliating the expandable graphite to obtain a compressible mixture comprising expanded graphite worms and the non-expandable component; (e) compressing the compressible mixture at a pressure within the range of from about 5 psi to about 50,000 psi in predetermined directions into predetermined forms of cohered graphite composite compact; and (f) treating the so-formed cohered graphite composite to activate the binder or matrix material thereby promoting adhesion within the compact to produce the desired composite composition. Preferably, the non-expandable powder component further comprises an isotropy-promoting agent such as non-expandable graphite particles. Further preferably, step (e) comprises compressing the mixture in at least two directions. The method leads to composite plates with exceptionally high thickness-direction electrical conductivity.

  18. Fuel System: Automotive Mechanics Instructional Program. Block 4.

    Science.gov (United States)

    O'Brien, Ralph D.

    The fourth of six instructional blocks in automotive mechanics, the lessons and supportive information in the document provide a guide for teachers in planning an instructional program in automotive fuel systems at the secondary and post secondary level. The material, as organized, is a suggested sequence of instruction within each block. Each…

  19. Method to fabricate block fuel elements for high temperature reactors

    International Nuclear Information System (INIS)

    Hrovat, M.; Rachor, L.

    1977-01-01

    The fabrication of block fuel elements for gas-cooled high temperature reactors can be improved upon by adding 0.2 to 2 wt.% of a hydrocarbon compound to the lubricating mixture prior to pressing. Hexanol or octanol are named as substances. The dimensional accuracy of the block is thus improved. 2 examples illustrate the method. (RW) [de

  20. Method to fabricate block fuel elements for high temperature reactors

    International Nuclear Information System (INIS)

    Hrovat, M.; Rachor, L.

    1978-01-01

    The fabrication of block fuel elements for gas-cooled high temperature reactors can be improved upon by adding 0.2 to 2 wt.% of a hydrocarbon compound to the lubricating mixture prior to pressing. Hexanol or octanol are named as substances. The dimensional accuracy of the block is thus improved. 2 examples illustrate the method. (orig./PW)

  1. A Comparison of Materials Issues for Cermet and Graphite-Based NTP Fuels

    Science.gov (United States)

    Stewart, Mark E.; Schnitzler, Bruce G.

    2013-01-01

    This paper compares material issues for cermet and graphite fuel elements. In particular, two issues in NTP fuel element performance are considered here: ductile to brittle transition in relation to crack propagation, and orificing individual coolant channels in fuel elements. Their relevance to fuel element performance is supported by considering material properties, experimental data, and results from multidisciplinary fluid/thermal/structural simulations. Ductile to brittle transition results in a fuel element region prone to brittle fracture under stress, while outside this region, stresses lead to deformation and resilience under stress. Poor coolant distribution between fuel element channels can increase stresses in certain channels. NERVA fuel element experimental results are consistent with this interpretation. An understanding of these mechanisms will help interpret fuel element testing results.

  2. Multidisciplinary Simulation of Graphite-Composite and Cermet Fuel Elements for NTP Point of Departure Designs

    Science.gov (United States)

    Stewart, Mark E.; Schnitzler, Bruce G.

    2015-01-01

    This paper compares the expected performance of two Nuclear Thermal Propulsion fuel types. High fidelity, fluid/thermal/structural + neutronic simulations help predict the performance of graphite-composite and cermet fuel types from point of departure engine designs from the Nuclear Thermal Propulsion project. Materials and nuclear reactivity issues are reviewed for each fuel type. Thermal/structural simulations predict thermal stresses in the fuel and thermal expansion mis-match stresses in the coatings. Fluid/thermal/structural/neutronic simulations provide predictions for full fuel elements. Although NTP engines will utilize many existing chemical engine components and technologies, nuclear fuel elements are a less developed engine component and introduce design uncertainty. Consequently, these fuel element simulations provide important insights into NTP engine performance.

  3. 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

  4. Non-catalyzed cathodic oxygen reduction at graphite granules in microbial fuel cells

    International Nuclear Information System (INIS)

    Freguia, Stefano; Rabaey, Korneel; Yuan Zhiguo; Keller, Juerg

    2007-01-01

    Oxygen is the most sustainable electron acceptor currently available for microbial fuel cell (MFC) cathodes. However, its high overpotential for reduction to water limits the current that can be produced. Several materials and catalysts have previously been investigated in order to facilitate oxygen reduction at the cathode surface. This study shows that significant stable currents can be delivered by using a non-catalyzed cathode made of granular graphite. Power outputs up to 21 W m -3 (cathode total volume) or 50 W m -3 (cathode liquid volume) were attained in a continuous MFC fed with acetate. These values are higher than those obtained in several other studies using catalyzed graphite in various forms. The presence of nanoscale pores on granular graphite provides a high surface area for oxygen reduction. The current generated with this cathode can sustain an anodic volume specific COD removal rate of 1.46 kg COD m -3 d -1 , which is higher than that of a conventional aerobic process. This study demonstrates that microbial fuel cells can be operated efficiently using high surface graphite as cathode material. This implies that research on microbial fuel cell cathodes should not only focus on catalysts, but also on high surface area materials

  5. Non-catalyzed cathodic oxygen reduction at graphite granules in microbial fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Freguia, Stefano; Rabaey, Korneel; Yuan, Zhiguo; Keller, Juerg [The University of Queensland, St. Lucia, Qld (Australia). Advanced Wastewater Management Centre

    2007-12-01

    Oxygen is the most sustainable electron acceptor currently available for microbial fuel cell (MFC) cathodes. However, its high overpotential for reduction to water limits the current that can be produced. Several materials and catalysts have previously been investigated in order to facilitate oxygen reduction at the cathode surface. This study shows that significant stable currents can be delivered by using a non-catalyzed cathode made of granular graphite. Power outputs up to 21 W m{sup -3} (cathode total volume) or 50 W m{sup -3} (cathode liquid volume) were attained in a continuous MFC fed with acetate. These values are higher than those obtained in several other studies using catalyzed graphite in various forms. The presence of nanoscale pores on granular graphite provides a high surface area for oxygen reduction. The current generated with this cathode can sustain an anodic volume specific COD removal rate of 1.46 kg{sub COD} m{sup -3} d{sup -1}, which is higher than that of a conventional aerobic process. This study demonstrates that microbial fuel cells can be operated efficiently using high surface graphite as cathode material. This implies that research on microbial fuel cell cathodes should not only focus on catalysts, but also on high surface area materials. (author)

  6. Anisotropic Material Behavior of Uni-axially Compacted Graphite Matrix for HTGR Fuel Compact Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Yoon, Ji-Hae; Cho, Moon Sung [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In developing the fuel compact fabrication technology, and fuel graphite material to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions and the material properties of fuel element. It was observed, during this development, that the pressing technique employed for the compaction fabrication prior to the two successive heat treatments (carbonization and final high temperature heat treatment) was of extreme importance in determining the material properties of the final compact product. In this work, the material behavior of the uni-axially pressed graphite matrix during the carbonization and final heat treatment are evaluated and summarized along the different directions, viz., perpendicular and parallel directions to pressing direction. In this work, the dimensional variations and variations in thermal expansion, thermal conductivity and Vickers hardness of the graphite matrix compact samples in the axial and radial directions prepared by uni-axial pressing are evaluated, and compared with those of samples prepared by cold isostatic pressing with the available data. From this work, the followings are observed. 1) Dimensional changes of matrix graphite green compacts during carbonization show that the difference in radial and axial variations shows a large anisotropic behavior in shrinkage. The radial variation is very small while the axial variation is large. During carbonization, the stresses caused by the force would be released in to the axial direction together with the phenolic resin vapor. 2) Dimensional variation of compact samples in perpendicular and parallel directions during carbonization shows a large difference in behavior when compact sample is prepared by uni-axial pressing. However, when compact sample is prepared by cold isostatic pressing, there is

  7. Anisotropic Material Behavior of Uni-axially Compacted Graphite Matrix for HTGR Fuel Compact Fabrication

    International Nuclear Information System (INIS)

    Lee, Young-Woo; Yeo, Seunghwan; Yoon, Ji-Hae; Cho, Moon Sung

    2016-01-01

    In developing the fuel compact fabrication technology, and fuel graphite material to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions and the material properties of fuel element. It was observed, during this development, that the pressing technique employed for the compaction fabrication prior to the two successive heat treatments (carbonization and final high temperature heat treatment) was of extreme importance in determining the material properties of the final compact product. In this work, the material behavior of the uni-axially pressed graphite matrix during the carbonization and final heat treatment are evaluated and summarized along the different directions, viz., perpendicular and parallel directions to pressing direction. In this work, the dimensional variations and variations in thermal expansion, thermal conductivity and Vickers hardness of the graphite matrix compact samples in the axial and radial directions prepared by uni-axial pressing are evaluated, and compared with those of samples prepared by cold isostatic pressing with the available data. From this work, the followings are observed. 1) Dimensional changes of matrix graphite green compacts during carbonization show that the difference in radial and axial variations shows a large anisotropic behavior in shrinkage. The radial variation is very small while the axial variation is large. During carbonization, the stresses caused by the force would be released in to the axial direction together with the phenolic resin vapor. 2) Dimensional variation of compact samples in perpendicular and parallel directions during carbonization shows a large difference in behavior when compact sample is prepared by uni-axial pressing. However, when compact sample is prepared by cold isostatic pressing, there is

  8. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    International Nuclear Information System (INIS)

    Karahan, Aydın; Kazimi, Mujid S.

    2013-01-01

    The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors

  9. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydın, E-mail: karahan@alum.mit.edu; Kazimi, Mujid S.

    2013-10-15

    The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

  10. Electroanalysis of cardioselective beta-adrenoreceptor blocking agent acebutolol by disposable graphite pencil electrodes with detailed redox mechanism

    OpenAIRE

    Atmanand M. Bagoji; Shreekant M. Patil; Sharanappa T. Nandibewoor

    2016-01-01

    A simple economic graphite pencil electrode (GPE) was used for analysis of cardioselective, hydrophilic-adrenoreceptor blocking agent, acebutolol (ACBT) using the cyclic voltammetric, linear sweep voltammetric, differential pulse voltammetric (DPV), and square-wave voltammetric (SWV) techniques. The dependence of the current on pH, concentration, and scan rate was investigated to optimize the experimental condition for determination of ACBT. The electrochemical behavior of the ACBT at GPE was...

  11. Study on thermal conductivity of HTR spherical fuel element matrix graphite

    International Nuclear Information System (INIS)

    Zhang Kaihong; Liu Xiaoxue; Zhao Hongsheng; Li Ziqiang; Tang Chunhe

    2014-01-01

    Taking the spherical fuel element matrix graphite ball samples as an example, this paper introduced the principle and method of laser thermal conductivity meter, as well as the specific heat capacity, and analyzed the effects of different test methods and sampling methods on the thermal conductivities at 1000 ℃ of graphite material. The experimental results show that the thermal conductivities of graphite materials tested by synchronous thermal analyzer combining with laser thermal conductivity meter were different from that directly by laser thermal conductivity meter, the former was more reliable and accurate than the later; When sampling from different positions, central samples had higher thermal conductivities than edging samples, which was related to the material density and porosity at the different locations; the thermal conductivities had obvious distinction between samples from different directions, which was because the layer structure of polycrystalline graphite preferred orientation under pressure, generally speaking, the thermal conductivities perpendicular to the molding direction were higher than that parallel to the molding direction. Besides this, the test results show that the thermal conductivities of all the graphite material samples were greater than 30 W/(m (K), achieving the thermal performance index of high temperature gas cooled reactor. (authors)

  12. Dimensional Behavior of Matrix Graphite Compacts during Heat Treatments for HTGR Fuel Element Fabrication

    International Nuclear Information System (INIS)

    Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung

    2015-01-01

    The carbonization is a process step where the binder that is incorporated during the matrix graphite powder preparation step is evaporated and the residue of the binder is carbonized during the heat treatment at about 1073 K. This carbonization step is followed by the final high temperature heat treatment where the carbonized compacts are heat treated at 2073-2173 K in vacuum for a relatively short time (about 2 hrs). In order to develop a fuel compact fabrication technology, and for fuel matrix graphite to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions, which has a strong influence on the further steps and the material properties of fuel element. In this work, the dimensional changes of green compacts during the carbonization and final heat treatment are evaluated when compacts have different densities from different pressing conditions and different final heat treatment temperatures are employed, keeping other process parameters constant, such as the binder content, carbonization time, temperature and atmosphere (two hours ant 1073K and N2 atmosphere). In this work, the dimensional variations of green compacts during the carbonization and final heat treatment are evaluated when compacts have different densities from different pressing conditions and different final heat treatment temperatures are employed

  13. 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.

  14. Method to produce fuel element blocks for HTR reactors

    International Nuclear Information System (INIS)

    Hrovat, M.; Rachor, L.

    1977-01-01

    The patent claim relates to one partial step of the multi-stage pressing process in the production of fuel elements. A binder resin with a softening point at least 15 0 C but preferably 25-40 0 C above the melting point of the lubricant is proposed. The pressed block is expelled from the forging die in the temperature interval between the melting point of the lubricant and the softening point of the binder resin. The purpose of the invention is that the pressed fuel element blocks are expelled from the machine tool without damage at a pressure low enough to protect the mechanical integrity of the coated fuel particles or fertile particles. (UA) [de

  15. PEM fuel cells with injection moulded bipolar plates of highly filled graphite compounds; PEM-Brennstoffzellen mit spritzgegossenen Bipolarplatten aus hochgefuelltem Graphit-Compound

    Energy Technology Data Exchange (ETDEWEB)

    Kreuz, Can

    2008-04-11

    This work concerns with the injection moulding of highly filled graphite compounds to bipolar plates for PEM fuel cells in a power output range between 100 - 500 Watts. A particular focus is laid on the combination of the three multidisciplinary scopes like material development, production technology and component development / design. The results of the work are specified by the process-oriented characterisation of the developed and manufactured bipolar plates as well as their application in a functioning fuel cell. (orig.)

  16. Solid fuel block as an alternate fuel for cooking and barbecuing: Preliminary results

    International Nuclear Information System (INIS)

    Sharma, Monikankana; Mukunda, H.S.; Sridhar, G.

    2009-01-01

    A large part of the rural people of developing countries use traditional biomass stoves to meet their cooking and heating energy demands. These stoves possess very low thermal efficiency; besides, most of them cannot handle agricultural wastes. Thus, there is a need to develop an alternate cooking contrivance which is simple, efficient and can handle a range of biomass including agricultural wastes. In this reported work, a highly densified solid fuel block using a range of low cost agro residues has been developed to meet the cooking and heating needs. A strategy was adopted to determine the best suitable raw materials, which was optimized in terms of cost and performance. Several experiments were conducted using solid fuel block which was manufactured using various raw materials in different proportions; it was found that fuel block composed of 40% biomass, 40% charcoal powder, 15% binder and 5% oxidizer fulfilled the requirement. Based on this finding, fuel blocks of two different configurations viz. cylindrical shape with single and multi-holes (3, 6, 9 and 13) were constructed and its performance was evaluated. For instance, the 13 hole solid fuel block met the requirement of domestic cooking; the mean thermal power was 1.6 kW th with a burn time of 1.5 h. Furthermore, the maximum thermal efficiency recorded for this particular design was 58%. Whereas, the power level of single hole solid fuel block was found to be lower but adequate for barbecue cooking application

  17. Process for the production of fuel combined articles for addition in block shaped high temperature fuel elements

    International Nuclear Information System (INIS)

    Hrovat, M.; Rachor, L.

    1976-01-01

    There is provided a process for the production of fuel compacts consisting of an isotropic, radiation-resistant graphite matrix of good heat conductivity having embedded therein coated fuel and/or fertile particles for insertion into high temperature fuel elements by providing the coated fuel and/or fertile particles with an overcoat of molding mixture consisting of graphite powder and a thermoplastic resin binder. The particles after the overcoating are provided with hardener and lubricant only on the surface and subsequently are compressed in a die heated to a constant temperature of about 150 0 C, hardened and discharged therefrom as finished compacts

  18. 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

  19. Laminated exfoliated graphite composite-metal compositions for fuel cell flow field plate or bipolar plate applications

    Science.gov (United States)

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

    2014-05-20

    An electrically conductive laminate composition for fuel cell flow field plate or bipolar plate applications. The laminate composition comprises at least a thin metal sheet having two opposed exterior surfaces and a first exfoliated graphite composite sheet bonded to the first of the two exterior surfaces of the metal sheet wherein the exfoliated graphite composite sheet comprises: (a) expanded or exfoliated graphite and (b) a binder or matrix material to bond the expanded graphite for forming a cohered sheet, wherein the binder or matrix material is between 3% and 60% by weight based on the total weight of the first exfoliated graphite composite sheet. Preferably, the first exfoliated graphite composite sheet further comprises particles of non-expandable graphite or carbon in the amount of between 3% and 60% by weight based on the total weight of the non-expandable particles and the expanded graphite. Further preferably, the laminate comprises a second exfoliated graphite composite sheet bonded to the second surface of the metal sheet to form a three-layer laminate. Surface flow channels and other desired geometric features can be built onto the exterior surfaces of the laminate to form a flow field plate or bipolar plate. The resulting laminate has an exceptionally high thickness-direction conductivity and excellent resistance to gas permeation.

  20. A Study on Field Emission Characteristics of Planar Graphene Layers Obtained from a Highly Oriented Pyrolyzed Graphite Block.

    KAUST Repository

    Lee, Seok Woo; Lee, Seung S; Yang, Eui-Hyeok

    2009-01-01

    This paper describes an experimental study on field emission characteristics of individual graphene layers for vacuum nanoelectronics. Graphene layers were prepared by mechanical exfoliation from a highly oriented pyrolyzed graphite block and placed on an insulating substrate, with the resulting field emission behavior investigated using a nanomanipulator operating inside a scanning electron microscope. A pair of tungsten tips controlled by the nanomanipulator enabled electric connection with the graphene layers without postfabrication. The maximum emitted current from the graphene layers was 170 nA and the turn-on voltage was 12.1 V.

  1. A Study on Field Emission Characteristics of Planar Graphene Layers Obtained from a Highly Oriented Pyrolyzed Graphite Block.

    KAUST Repository

    Lee, Seok Woo

    2009-07-12

    This paper describes an experimental study on field emission characteristics of individual graphene layers for vacuum nanoelectronics. Graphene layers were prepared by mechanical exfoliation from a highly oriented pyrolyzed graphite block and placed on an insulating substrate, with the resulting field emission behavior investigated using a nanomanipulator operating inside a scanning electron microscope. A pair of tungsten tips controlled by the nanomanipulator enabled electric connection with the graphene layers without postfabrication. The maximum emitted current from the graphene layers was 170 nA and the turn-on voltage was 12.1 V.

  2. Graphitic Carbon Nitride as a Catalyst Support in Fuel Cells and Electrolyzers

    International Nuclear Information System (INIS)

    Mansor, Noramalina; Miller, Thomas S.; Dedigama, Ishanka; Jorge, Ana Belen; Jia, Jingjing; Brázdová, Veronika; Mattevi, Cecilia; Gibbs, Chris; Hodgson, David; Shearing, Paul R.; Howard, Christopher A.; Corà, Furio; Shaffer, Milo; Brett, Daniel J.L.

    2016-01-01

    Highlights: • Graphitic carbon nitride (gCN) describes many materials with different structures. • gCNs can exhibit excellent mechanical, chemical and thermal resistance. • A major obstacle for pure gCN catalyst supports is limited electronic conductivity. • Composite/Hybrid gCN structures show excellent performance as catalyst supports. • gCNs have great potential for use in fuel calls and water electrolyzers. - Abstract: Electrochemical power sources, such as polymer electrolyte membrane fuel cells (PEMFCs), require the use of precious metal catalysts which are deposited as nanoparticles onto supports in order to minimize their mass loading and therefore cost. State-of-the-art/commercial supports are based on forms of carbon black. However, carbon supports present disadvantages including corrosion in the operating fuel cell environment and loss of catalyst activity. Here we review recent work examining the potential of different varieties of graphitic carbon nitride (gCN) as catalyst supports, highlighting their likely benefits, as well as the challenges associated with their implementation. The performance of gCN and hybrid gCN-carbon materials as PEMFC electrodes is discussed, as well as their potential for use in alkaline systems and water electrolyzers. We illustrate the discussion with examples taken from our own recent studies.

  3. In situ formation of graphene layers on graphite surfaces for efficient anodes of microbial fuel cells.

    Science.gov (United States)

    Tang, Jiahuan; Chen, Shanshan; Yuan, Yong; Cai, Xixi; Zhou, Shungui

    2015-09-15

    Graphene can be used to improve the performance of the anode in a microbial fuel cell (MFC) due to its good biocompatibility, high electrical conductivity and large surface area. However, the chemical production and modification of the graphene on the anode are environmentally hazardous because of the use of various harmful chemicals. This study reports a novel method based on the electrochemical exfoliation of a graphite plate (GP) for the in situ formation of graphene layers on the surface of a graphite electrode. When the resultant graphene-layer-based graphite plate electrode (GL/GP) was used as an anode in an MFC, a maximum power density of 0.67 ± 0.034 W/m(2) was achieved. This value corresponds to 1.72-, 1.56- and 1.26-times the maximum power densities of the original GP, exfoliated-graphene-modified GP (EG/GP) and chemically-reduced-graphene-modified GP (rGO/GP) anodes, respectively. Electrochemical measurements revealed that the high performance of the GL/GP anode was attributable to its macroporous structure, improved electron transfer and high electrochemical capacitance. The results demonstrated that the proposed method is a facile and environmentally friendly synthesis technique for the fabrication of high-performance graphene-based electrodes for use in microbial energy harvesting. Copyright © 2015 Elsevier B.V. All rights reserved.

  4. Oxygen reduction kinetics on graphite cathodes in sediment microbial fuel cells.

    Science.gov (United States)

    Renslow, Ryan; Donovan, Conrad; Shim, Matthew; Babauta, Jerome; Nannapaneni, Srilekha; Schenk, James; Beyenal, Haluk

    2011-12-28

    Sediment microbial fuel cells (SMFCs) have been used as renewable power sources for sensors in fresh and ocean waters. Organic compounds at the anode drive anodic reactions, while oxygen drives cathodic reactions. An understanding of oxygen reduction kinetics and the factors that determine graphite cathode performance is needed to predict cathodic current and potential losses, and eventually to estimate the power production of SMFCs. Our goals were to (1) experimentally quantify the dependence of oxygen reduction kinetics on temperature, electrode potential, and dissolved oxygen concentration for the graphite cathodes of SMFCs and (2) develop a mechanistic model. To accomplish this, we monitored current on polarized cathodes in river and ocean SMFCs. We found that (1) after oxygen reduction is initiated, the current density is linearly dependent on polarization potential for both SMFC types; (2) current density magnitude increases linearly with temperature in river SMFCs but remains constant with temperature in ocean SMFCs; (3) the standard heterogeneous rate constant controls the current density temperature dependence; (4) river and ocean SMFC graphite cathodes have large potential losses, estimated by the model to be 470 mV and 614 mV, respectively; and (5) the electrochemical potential available at the cathode is the primary factor controlling reduction kinetic rates. The mechanistic model based on thermodynamic and electrochemical principles successfully fit and predicted the data. The data, experimental system, and model can be used in future studies to guide SMFC design and deployment, assess SMFC current production, test cathode material performance, and predict cathode contamination.

  5. A study on amphiphilic fluorinated block copolymer in graphite exfoliation using supercritical CO2 for stable graphene dispersion.

    Science.gov (United States)

    Kim, Young Hyun; Lee, Hyang Moo; Choi, Sung Wook; Cheong, In Woo

    2018-01-15

    In this study, poly(2,2,2-trifluoroethyl methacrylate)-block-poly(4-vinylpyridine) (PTFEMA-b-PVP) was synthesized by stepwise reversible addition-fragmentation chain transfer (RAFT) polymerization for the preparation of graphene by the exfoliation of graphite nanoplatelets (GPs) in supercritical CO 2 (SCCO 2 ). Two different block copolymers (low and high molecular weights) were prepared with the same block ratio and used at different concentrations in the SCCO 2 process. The amount of PTFEMA-b-PVP adsorbed on the GPs and the electrical conductivity of the SCCO 2 -treated GP samples were evaluated using thermogravimetric analysis (TGA) and four-point probe method, respectively. All GP samples treated with SCCO 2 were then dispersed in methanol and the dispersion stability was investigated using online turbidity measurements. The concentration and morphology of few-layer graphene stabilized with PTFEMA-b-PVP in the supernatant solution were investigated by gravimetry, scanning electron microscopy, and Raman spectroscopy. Destabilization study of the graphene dispersions revealed that the longer block copolymer exhibited better affinity for graphene, resulting in a higher yield of stable graphene with minimal defects. Copyright © 2017 Elsevier Inc. All rights reserved.

  6. Method of making a graphite fuel element having carbonaceous fuel bodies

    International Nuclear Information System (INIS)

    Miertschin, G.N.; Leary, D.F.

    1977-01-01

    Particulate nuclear fuel material, particulate carbon and pitch are combined with an additive which is effective to reduce the coke yield upon carbonization to mold a green fuel body. The additive may be polystyrene, a styrene-butadiene copolymer, an aromatic hydrocarbon having a molecular weight between about 75 and 300 or a saturated hydrocarbon polymer. The green fuel body is inserted in a complementary cavity within a porous nuclear fuel element body and heated in situ to decompose the pitch and additive, leaving a relatively close-fitting fuel body in the cavity

  7. 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

  8. 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.)

  9. The Performance of a Direct Borohydride/Peroxide Fuel Cell Using Graphite Felts as Electrodes

    Directory of Open Access Journals (Sweden)

    Heng-Yi Lee

    2017-08-01

    Full Text Available A direct borohydride/peroxide fuel cell (DBPFC generates electrical power by recirculating liquid anolyte and catholyte between the stack and reservoirs, which is similar to the operation of flow batteries. To enhance the accessibility of the catalyst layer to the liquid anolyte/catholyte, graphite felts are employed as the porous diffusion layer of a single-cell DBPFC instead of carbon paper/cloth. The effects of the type of anode alkaline solution and operating conditions, including flow rate and temperature of the anolyte/catholyte, on DBPFC performance are investigated and discussed. The durability of the DBPFC is also evaluated by galvanostatic discharge at 0.1 A∙cm−2 for over 50 h. The results of this preliminary study show that a DBPFC with porous graphite electrodes can provide a maximum power density of 0.24 W∙cm−2 at 0.8 V. The performance of the DBPFC drops slightly after 50 h of operation; however, the discharge capacity shows no significant decrease.

  10. A Graphite Oxide Paper Polymer Electrolyte for Direct Methanol Fuel Cells

    Directory of Open Access Journals (Sweden)

    Ravi Kumar

    2011-01-01

    Full Text Available A flow directed assembly of graphite oxide solution was used in the formation of free-standing graphene oxide paper of approximate thickness of 100 μm. The GO papers were characterised by XRD and SEM. Electrochemical characterization of the GO paper membrane electrode assembly revealed proton conductivities of 4.1 × 10−2 S cm−1 to 8.2 × 10−2 S cm−1 at temperatures of 25–90°C. A direct methanol fuel cell, at 60°C, gave a peak power density of 8 mW cm−2 at a current density of 35 mA cm−2.

  11. Properties of unirradiated fuel element graphites H-451 and SO818. [Bulk density, tensile properties, thermal expansion, thermal conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Engle, G.B.; Johnson, W.R.

    1976-10-08

    Nuclear graphites H-451, lot 440 (Great Lakes Carbon Corporation (GLCC)), and SO818 (Airco Speer Division, Air Reduction Corporation (AS)) are described, and physical, mechanical, and chemical property data are presented for the graphites in the unirradiated state. A summary of the mean values of the property data and of data on TS-1240 and H-451, lot 426, is tabulated. A direct comparison of H-451, lot 426, chosen for Fort St. Vrain (FSV) fuel reload production, TS-1240, and SO818 may be made from the table. (auth)

  12. Factors affecting defective fraction of biso-coated HTGR fuel particles during in-block carbonization

    International Nuclear Information System (INIS)

    Caputo, A.J.; Johnson, D.R.; Bayne, C.K.

    1977-01-01

    The performance of Biso-coated thoria fuel particles during the in-block processing step of HTGR fuel element refabrication was evaluated. The effect of various process variables (heating rate, particle crushing strength, horizontal and/or vertical position in the fuel element blocks, and fuel hole permeability) on pitch coke yield, defective fraction of fuel particles, matrix structure, and matrix porosity was evaluated. Of the variables tested, only heating rate had a significant effect on pitch coke yield while both heating rate and particle crushing strength had a significant effect on defective fraction of fuel particles

  13. Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Yajuan, E-mail: yajuan.zhong@gmail.com [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences, Taiyuan 030001 (China); Zhang, Junpeng [CAS Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences, Taiyuan 030001 (China); Lin, Jun, E-mail: linjun@sinap.ac.cn [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Xu, Liujun [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Guo, Quangui [CAS Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences, Taiyuan 030001 (China)

    2017-07-15

    Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10{sup −6} K{sup −1} (α{sub ∥}) and 6.15 × 10{sup −6} K{sup −1} (α{sub ⊥}) at the temperature range of 25–700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.

  14. Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

    International Nuclear Information System (INIS)

    Zhong, Yajuan; Zhang, Junpeng; Lin, Jun; Xu, Liujun; Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong; Guo, Quangui

    2017-01-01

    Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10 −6 K −1 (α ∥ ) and 6.15 × 10 −6 K −1 (α ⊥ ) at the temperature range of 25–700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.

  15. Milling Behavior of Matrix Graphite Powders with Different Binder Materials in HTGR Fuel Element Fabrication: I. Variation in Particle Size Distribution

    International Nuclear Information System (INIS)

    Lee, Young Woo; Cho, Moon Sung

    2011-01-01

    The fuel element for HTGR is manufactured by mixing coated fuel particles with matrix graphite powder and forming into either pebble type or cylindrical type compacts depending on their use in different HTGR cores. The coated fuel particle, the so-called TRISO particle, consists of 500-μm spherical UO 2 particles coated with the low density buffer Pyrolytic Carbon (PyC) layer, the inner and outer high density PyC layer and SiC layer sandwiched between the two inner and outer PyC layers. The coated TRISO particles are mixed with a matrix graphite powder properly prepared and pressed into a spherical shape or a cylindrical compact finally heat-treated at about 1900 .deg. C. These fuel elements can have different sizes and forms of compact. The basic steps for manufacturing a fuel element include preparation of graphite matrix powder, overcoating the fuel particles, mixing the fuel particles with a matrix powder, carbonizing green compact, and the final high-temperature heat treatment of the carbonized fuel compact. In order to develop a fuel compact fabrication technology, it is important to develop a technology to prepare the matrix graphite powder (MGP) with proper characteristics, which has a strong influence on further steps and the material properties of fuel element. In this work, the milling behavior of matrix graphite powder mixture with different binder materials and their contents was investigated by analyzing the change in particle size distribution with different milling time

  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. Electroanalysis of cardioselective beta-adrenoreceptor blocking agent acebutolol by disposable graphite pencil electrodes with detailed redox mechanism

    Directory of Open Access Journals (Sweden)

    Atmanand M. Bagoji

    2016-12-01

    Full Text Available A simple economic graphite pencil electrode (GPE was used for analysis of cardioselective, hydrophilic-adrenoreceptor blocking agent, acebutolol (ACBT using the cyclic voltammetric, linear sweep voltammetric, differential pulse voltammetric (DPV, and square-wave voltammetric (SWV techniques. The dependence of the current on pH, concentration, and scan rate was investigated to optimize the experimental condition for determination of ACBT. The electrochemical behavior of the ACBT at GPE was a diffusion-controlled process. A probable electro-redox mechanism was proposed. Under the optimal conditions, the anodic peak current was linearly proportional to the concentration of ACBT in the range from 1.00 to 15.0 μM with a limit of detection 1.26 × 10−8 M for DPV and 1.28 × 10−8 M for the SWV. This method was applied for quantitative determination of the ACBT levels in urine as real samples. The obtained recovery ranges for ACBT in urine were from 95.4 to101% as found by the standard addition technique. Further interference study was also carried with some common interfering substances.

  18. 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)

  19. Neutronic evaluation of a fuel block of a GT-MHR using WIMSD5

    International Nuclear Information System (INIS)

    Silva, Clarysson Alberto Mello da; Pereira, Claubia; Costa, Antonella Lombardi; Veloso, Maria Auxiliadora Fortini

    2009-01-01

    The goal is to simulate a representative fuel block of a GT-MHR core to analyze the neutronic parameters behavior due the insertion of Pu isotopes and Minor Actinides (MAs) using shuffling scheme. Initially the fuel block was filled with Driver Fuel (DF), and after burned, these fuels are reprocessed and build the Transmutation Fuel (TF). After some cycles, the fuel block was filled with DF and TF fuels. The DF is a mixture of Pu and Np and the TF is a mix of Pu and MAs. The shuffled scheme was evaluated after each cycle. It was verified that neutronic parameters and isotopic composition reach equilibrium and remain within safety limits when this scheme is used. In addition, there were burnup of MAs. The WIMS code was used in the simulations and the following neutronic parameters were evaluated: infinitive multiplication factor, spectrum hardening and reactivity temperature coefficients. (author)

  20. Structural characteristics of a graphite moderated critical assembly for a Zero Power reactor at IEA (Brazil)

    International Nuclear Information System (INIS)

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

    1975-01-01

    The structural characteristics of a graphite moderated core of a critical assembly to be installed in the Zero Power Reactor of IEA have been defined. These characteristics are the graphite block dimensions, the number and dimensions of the holes in the graphite, the pitch, the dimensions of the sticks of fuel and graphite to be inserted in the holes, and the mechanical reproducibility of the system. The composition of the fuel and moderator sticks were also defined. The main boundary conditions were the range of the relation C/U and C/TH used in commercial HTGR and the neutronics homogeneity

  1. Effect of NaX zeolite-modified graphite felts on hexavalent chromium removal in biocathode microbial fuel cells.

    Science.gov (United States)

    Wu, Xiayuan; Tong, Fei; Yong, Xiaoyu; Zhou, Jun; Zhang, Lixiong; Jia, Honghua; Wei, Ping

    2016-05-05

    Two kinds of NaX zeolite-modified graphite felts were used as biocathode electrodes in hexavalent chromium (Cr(VI))-reducing microbial fuel cells (MFCs). The one was fabricated through direct modification, and the other one processed by HNO3 pretreatment of graphite felt before modification. The results showed that two NaX zeolite-modified graphite felts are excellent bio-electrode materials for MFCs, and that a large NaX loading mass, obtained by HNO3 pretreatment (the HNO3-NaX electrode), leads to a superior performance. The HNO3-NaX electrode significantly improved the electricity generation and Cr(VI) removal of the MFC. The maximum Cr(VI) removal rate increased to 10.39±0.28 mg/L h, which was 8.2 times higher than that of the unmodified control. The improvement was ascribed to the strong affinity that NaX zeolite particles, present in large number on the graphite felt, have for microorganisms and Cr(VI) ions. Copyright © 2016 Elsevier B.V. All rights reserved.

  2. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  3. Block Copolymers for Alkaline Fuel Cell Membrane Materials

    Science.gov (United States)

    2014-07-30

    temperature fuel cells including proton exchange membrane fuel cell ( PEMFC ) and alkaline fuel cell (AFC) with operation temperature usually lower than 120...advantages over proton exchange membrane fuel cells ( PEMFCs ) resulting in the popularity of AFCs in the US space program.[8-11] The primary benefit AFC...offered over PEMFC is better electrochemical kinetics on the anode and cathode under the alkaline environment, which results in the ability to use

  4. Thorium utilisation in a small long-life HTR. Part III: Composite-rod fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Verrue, Jacques, E-mail: jacques.verrue@polytechnique.org [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); École Polytechnique (Member of ParisTech), 91128 Palaiseau Cedex (France); Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands)

    2014-02-15

    Highlights: • Composite-rod fuel blocks are proposed for a small block-type HTR. • An axial separation of fuel compacts is the most important feature. • Three patterns are presented to analyse the effects of the spatial distribution. • The spatial distribution has a large influence on the neutron spectrum. • Composite-rod fuel blocks reach a reactivity swing less than 4%. - Abstract: The U-Battery is a small long-life high temperature gas-cooled reactor (HTR) with power of 20 MWth. In order to increase its lifetime and diminish its reactivity swing, the concept of composite-rod fuel blocks with uranium and thorium was investigated. Composite-rod fuel blocks feature a specific axial separation between UO{sub 2} and ThO{sub 2} compacts in fuel rods. The design parameters, investigated by SCALE 6, include the number and spatial distribution of fuel compacts within the rods, the enrichment of uranium, the radii of fuel kernels and fuel compacts, and the packing fractions of uranium and thorium TRISO particles. The analysis shows that a lower moderation ratio and a larger inventory of heavy metals results in a lower reactivity swing. The optimal atomic carbon-to-heavy metal ratio depends on the mass fraction of U-235 and is commonly in the 160–200 range. The spatial distribution of the fuel compacts within the fuel rods has a large influence on the energy spectrum in each fuel compact and thus on the beginning-of-life reactivity and the reactivity swing. At end-of-life, the differences caused by the spatial distribution of the fuel compacts are smaller due to the fissions of U-233 in the ThO{sub 2} fuel compacts. This phenomenon enables to design fuel blocks with a very low reactivity swing, down to less than 4% in a 10-year lifetime. Among three types of thorium fuelled U-Battery blocks, the composite-rod fuel block achieves the highest end-of-life reactivity and the lowest reactivity swing.

  5. A comparison of integral block and tubular interacting fuel element concepts for low enrichment HTR

    Energy Technology Data Exchange (ETDEWEB)

    Desoisa, J A

    1972-04-15

    The tubular interacting fuel element has to date been the favoured U.K. high temperature reactor design. Recent attempts to lower fuel costs and the progress of the Fort St. Vrain reactor has focussed attention on alternative designs, and in particular on the attractive design simplicity of the integral block concept. The aim of this investigation is to compare the merits of both concepts from fuel cycle cost and thermal performance viewpoints and to determine whether optimization of the integral block concept leads to changes in the current design values of (a) fuel density, (b) Nc/Nu, and/or (c) mean discharge irradiation within the framework of present design limits.

  6. 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

  7. Organic-resistant screen-printed graphitic electrodes: Application to on-site monitoring of liquid fuels

    International Nuclear Information System (INIS)

    Almeida, Eduardo S.; Silva, Luiz A.J.; Sousa, Raquel M.F.; Richter, Eduardo M.; Foster, Christopher W.; Banks, Craig E.; Munoz, Rodrigo A.A.

    2016-01-01

    This work presents the potential application of organic-resistant screen-printed graphitic electrodes (SPGEs) for fuel analysis. The required analysis of the antioxidant 2,6-di-tert-butylphenol (2,6-DTBP) in biodiesel and jet fuel is demonstrated as a proof-of-concept. The screen-printing of graphite, Ag/AgCl and insulator inks on a polyester substrate (250 μm thickness) resulted in SPGEs highly compatible with liquid fuels. SPGEs were placed on a batch-injection analysis (BIA) cell, which was filled with a hydroethanolic solution containing 99% v/v ethanol and 0.1 mol L −1 HClO 4 (electrolyte). An electronic micropipette was connected to the cell to perform injections (100 μL) of sample or standard solutions. Over 200 injections can be injected continuously without replacing electrolyte and SPGE strip. Amperometric detection (+1.1 V vs. Ag/AgCl) of 2,6-DTBP provided fast (around 8 s) and precise (RSD = 0.7%, n = 12) determinations using an external calibration curve. The method was applied for the analysis of biodiesel and aviation jet fuel samples and comparable results with liquid and gas chromatographic analyses, typically required for biodiesel and jet fuel samples, were obtained. Hence, these SPGE strips are completely compatible with organic samples and their combination with the BIA cell shows great promise for routine and portable analysis of fuels and other organic liquid samples without requiring sophisticated sample treatments. - Highlights: • Organic-resistant screen-printed graphitic electrodes (SPGE) for (bio)fuels. • Screen-printing of conductive and insulator inks on thin polyester substrate. • Continuous detection of antioxidants in electrolyte with 99% v/v ethanol. • SPGE coupled with batch-injection analysis allows over 200 injections (100 μL). • Similar results to GC and HPLC analyses of biodiesel and aviation jet fuels.

  8. Organic-resistant screen-printed graphitic electrodes: Application to on-site monitoring of liquid fuels

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, Eduardo S.; Silva, Luiz A.J.; Sousa, Raquel M.F.; Richter, Eduardo M. [Universidade Federal de Uberlândia, Universidade Federal de Uberlândia, Av. João Naves de Ávila, 2121, Uberlândia, MG, 38408100 (Brazil); Foster, Christopher W.; Banks, Craig E. [Manchester Metropolitan University, Faculty of Science and the Environment, School of Science and the Environment, Division of Chemistry and Environmental Science, Manchester, M1 5GD, England (United Kingdom); Munoz, Rodrigo A.A., E-mail: raamunoz@iqufu.ufu.br [Universidade Federal de Uberlândia, Universidade Federal de Uberlândia, Av. João Naves de Ávila, 2121, Uberlândia, MG, 38408100 (Brazil)

    2016-08-31

    This work presents the potential application of organic-resistant screen-printed graphitic electrodes (SPGEs) for fuel analysis. The required analysis of the antioxidant 2,6-di-tert-butylphenol (2,6-DTBP) in biodiesel and jet fuel is demonstrated as a proof-of-concept. The screen-printing of graphite, Ag/AgCl and insulator inks on a polyester substrate (250 μm thickness) resulted in SPGEs highly compatible with liquid fuels. SPGEs were placed on a batch-injection analysis (BIA) cell, which was filled with a hydroethanolic solution containing 99% v/v ethanol and 0.1 mol L{sup −1} HClO{sub 4} (electrolyte). An electronic micropipette was connected to the cell to perform injections (100 μL) of sample or standard solutions. Over 200 injections can be injected continuously without replacing electrolyte and SPGE strip. Amperometric detection (+1.1 V vs. Ag/AgCl) of 2,6-DTBP provided fast (around 8 s) and precise (RSD = 0.7%, n = 12) determinations using an external calibration curve. The method was applied for the analysis of biodiesel and aviation jet fuel samples and comparable results with liquid and gas chromatographic analyses, typically required for biodiesel and jet fuel samples, were obtained. Hence, these SPGE strips are completely compatible with organic samples and their combination with the BIA cell shows great promise for routine and portable analysis of fuels and other organic liquid samples without requiring sophisticated sample treatments. - Highlights: • Organic-resistant screen-printed graphitic electrodes (SPGE) for (bio)fuels. • Screen-printing of conductive and insulator inks on thin polyester substrate. • Continuous detection of antioxidants in electrolyte with 99% v/v ethanol. • SPGE coupled with batch-injection analysis allows over 200 injections (100 μL). • Similar results to GC and HPLC analyses of biodiesel and aviation jet fuels.

  9. Organic-resistant screen-printed graphitic electrodes: Application to on-site monitoring of liquid fuels.

    Science.gov (United States)

    Almeida, Eduardo S; Silva, Luiz A J; Sousa, Raquel M F; Richter, Eduardo M; Foster, Christopher W; Banks, Craig E; Munoz, Rodrigo A A

    2016-08-31

    This work presents the potential application of organic-resistant screen-printed graphitic electrodes (SPGEs) for fuel analysis. The required analysis of the antioxidant 2,6-di-tert-butylphenol (2,6-DTBP) in biodiesel and jet fuel is demonstrated as a proof-of-concept. The screen-printing of graphite, Ag/AgCl and insulator inks on a polyester substrate (250 μm thickness) resulted in SPGEs highly compatible with liquid fuels. SPGEs were placed on a batch-injection analysis (BIA) cell, which was filled with a hydroethanolic solution containing 99% v/v ethanol and 0.1 mol L(-1) HClO4 (electrolyte). An electronic micropipette was connected to the cell to perform injections (100 μL) of sample or standard solutions. Over 200 injections can be injected continuously without replacing electrolyte and SPGE strip. Amperometric detection (+1.1 V vs. Ag/AgCl) of 2,6-DTBP provided fast (around 8 s) and precise (RSD = 0.7%, n = 12) determinations using an external calibration curve. The method was applied for the analysis of biodiesel and aviation jet fuel samples and comparable results with liquid and gas chromatographic analyses, typically required for biodiesel and jet fuel samples, were obtained. Hence, these SPGE strips are completely compatible with organic samples and their combination with the BIA cell shows great promise for routine and portable analysis of fuels and other organic liquid samples without requiring sophisticated sample treatments. Copyright © 2016 Elsevier B.V. All rights reserved.

  10. An explication of the Graphite Structural Design Code of core components for the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Iyoku, Tatsuo; Ishihara, Masahiro; Toyota, Junji; Shiozawa, Shusaku

    1991-05-01

    The integrity evaluation of the core graphite components for the High Temperature Engineering Test Reactor (HTTR) will be carried out based upon the Graphite Structural Design Code for core components. In the application of this design code, it is necessary to make clear the basic concept to evaluate the integrity of core components of HTTR. Therefore, considering the detailed design of core graphite structures such as fuel graphite blocks, etc. of HTTR, this report explicates the design code in detail about the concepts of stress and fatigue limits, integrity evaluation method of oxidized graphite components and thermal irradiation stress analysis method etc. (author)

  11. 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)

  12. Graphite anode surface modification with controlled reduction of specific aryl diazonium salts for improved microbial fuel cells power output.

    Science.gov (United States)

    Picot, Matthieu; Lapinsonnière, Laure; Rothballer, Michael; Barrière, Frédéric

    2011-10-15

    Graphite electrodes were modified with reduction of aryl diazonium salts and implemented as anodes in microbial fuel cells. First, reduction of 4-aminophenyl diazonium is considered using increased coulombic charge density from 16.5 to 200 mC/cm(2). This procedure introduced aryl amine functionalities at the surface which are neutral at neutral pH. These electrodes were implemented as anodes in "H" type microbial fuel cells inoculated with waste water, acetate as the substrate and using ferricyanide reduction at the cathode and a 1000 Ω external resistance. When the microbial anode had developed, the performances of the microbial fuel cells were measured under acetate saturation conditions and compared with those of control microbial fuel cells having an unmodified graphite anode. We found that the maximum power density of microbial fuel cell first increased as a function of the extent of modification, reaching an optimum after which it decreased for higher degree of surface modification, becoming even less performing than the control microbial fuel cell. Then, the effect of the introduction of charged groups at the surface was investigated at a low degree of surface modification. It was found that negatively charged groups at the surface (carboxylate) decreased microbial fuel cell power output while the introduction of positively charged groups doubled the power output. Scanning electron microscopy revealed that the microbial anode modified with positively charged groups was covered by a dense and homogeneous biofilm. Fluorescence in situ hybridization analyses showed that this biofilm consisted to a large extent of bacteria from the known electroactive Geobacter genus. In summary, the extent of modification of the anode was found to be critical for the microbial fuel cell performance. The nature of the chemical group introduced at the electrode surface was also found to significantly affect the performance of the microbial fuel cells. The method used for

  13. Thorium utilization in a small long-life HTR. Part I: Th/U MOX fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands)

    2014-02-15

    Highlights: • We propose thorium MOX (TMOX) fuel blocks for a small block-type HTR. • The TMOX fuel blocks with low-enriched uranium are recommended. • More thorium decreases the reactivity swing of the TMOX fuel blocks. • Thorium reduces the negative temperature coefficient of the TMOX fuel blocks. • Thorium increases the conversion ratio of the TMOX fuel blocks. - Abstract: The U-Battery is a small, long-life and transportable high temperature gas-cooled reactor (HTR). The neutronic features of a typical fuel block with uranium and thorium have been investigated for a application of the U-Battery, by parametrically analyzing the composition and geometric parameters. The type of fuel block is defined as Th/U MOX fuel block because uranium and thorium are assumed to be mixed in each fuel kernel as a form of (Th,U)O{sub 2}. If the initially loaded mass of U-235 is mostly consumed in the early period of the lifetime of Th/U MOX fuel block, low-enriched uranium (LEU) as ignited fuel will not largely reduce the neutronic performance of the Th/U MOX fuel block, compared with high-enriched uranium. The radii of fuel kernels and fuel compacts and packing fraction of TRISO particles determine the atomic ratio of the carbon to heavy metal. When the ratio is smaller than 400, the difference among them due to double heterogeneous effects can be neglected for the Th/U MOX fuel block. In the range between 200 and 400, the reactivity swing of the Th/U MOX fuel block during 10 years is sufficiently small. The magnitude of the negative reactivity temperature coefficients of the Th/U MOX fuel block decreases by 20–45%, which is positive to reduce temperature defect of the Th/U MOX fuel block. The conversion ratio (CR) of the fuel increases from 0.48 (typical CR of the LEU-fueled U-Battery) to 0.78. The larger conversion ratio of the Th/U MOX fuel block reduces the reactivity swing during 10 years for the U-Battery.

  14. Recent developments in graphite. [Use in HTGR and aerospace

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Spontaneous modification of graphite anode by anthraquinone-2-sulfonic acid for microbial fuel cells.

    Science.gov (United States)

    Tang, Xinhua; Li, Haoran; Du, Zhuwei; Ng, How Yong

    2014-07-01

    In this study, anthraquinone-2-sulfonic acid (AQS), an electron transfer mediator, was immobilized onto graphite felt surface via spontaneous reduction of the in situ generated AQS diazonium cations. Cyclic voltammetry (CV) and energy dispersive spectrometry (EDS) characterizations of AQS modified graphite demonstrated that AQS was covalently grafted onto the graphite surface. The modified graphite, with a surface AQS concentration of 5.37 ± 1.15 × 10(-9)mol/cm(2), exhibited good electrochemical activity and high stability. The midpoint potential of the modified graphite was about -0.248 V (vs. normal hydrogen electrode, NHE), indicating that electrons could be easily transferred from NADH in bacteria to the electrode. AQS modified anode in MFCs increased the maximum power density from 967 ± 33 mW/m(2) to 1872 ± 42 mW/m(2). These results demonstrated that covalently modified AQS functioned as an electron transfer mediator to facilitate extracellular electron transfer from bacteria to electrode and significantly enhanced the power production in MFCs. Copyright © 2014 Elsevier Ltd. All rights reserved.

  16. Thorium utilization in a small long-life HTR. Part II: Seed-and-blanket fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Ming, E-mail: dingming@hrbeu.edu.cn [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands)

    2014-02-15

    Highlights: • Seed-and-blanket (S and B) fuel blocks are proposed for a small block-type HTR. • S and B fuel blocks consist of a seed region (UO{sub 2}) and a blanket region (ThO{sub 2}). • The neutronic performance of S and B fuel blocks are analyzed using SCALE 6. • Three S and B fuel blocks with a reactivity swing of 0.1 Δk are recommended. • S and B fuel blocks are compared with thorium MOX fuel blocks. - Abstract: In order to utilize thorium in high temperature gas-cooled reactors (HTRs), the concept of seed-and-blanket (S and B) fuel block is introduced into the U-Battery, which is a long-life block-type HTR with a thermal power of 20 MWth. A S and B fuel block consists of a seed region with uranium in the center, and a blanket region with thorium. The neutronic performance, such as the multiplication factor, conversion ratio and reactivity swing, of a typical S and B fuel block was investigated by SCALE 6.0 by parametric analysis of the composition parameters and geometric parameters of the fuel block for the U-Battery application. Since the purpose of U-235 in the S and B fuel block is to ignite the fission reactions in the fuel block, 20% enriched uranium is recommended for the S and B fuel block. When the ratio of the number of carbon to heavy metal atoms changes with the geometric parameters of the fuel block in the range of 200–250, the reactivity swing reaches very small values. Furthermore, for a reactivity swing of 0.1 Δk during 10 effective full power years, three configurations with 36, 54 and 78 UO{sub 2} fuel rods are recommended for the application of the U-Battery. The comparison analysis of the S and B fuel block with the Th/U MOX fuel block shows that the former has a longer lifetime and a lower reactivity swing.

  17. Nanostructuration of self-assembled poly(styrene-b-isoprene-b-styrene) block copolymer thin films in a highly oriented pyrolytic graphite substrate

    Energy Technology Data Exchange (ETDEWEB)

    Zalakain, Inaki; Ramos, Jose Angel; Fernandez, Raquel; Etxeberria, Haritz; Mondragon, Inaki, E-mail: inaki.mondragon@ehu.e

    2011-01-03

    Highly oriented pyrolitic graphite (HOPG) is a useful substrate to visualize epitaxial formation due to its crystallographic structure. The morphology of a poly(styrene-b-isoprene-b-styrene) block copolymer thin film on a HOPG substrate was investigated by atomic force microscopy. Block copolymer domains generated a morphology with triangular regularity. This arrangement was induced by the HOPG substrate structure due to van der Waals attraction between the HOPG {pi}-conjugated system and aromatic ring of polystyrene domains. However, increasing the film thickness, the substrate effect on the surface morphology decreased. As a consequence, film surfaces showed the coexistence of different structures such as highly aligned cylinders and perforated lamellae. When film thickness exceeded a threshold value, the substrate did not have effect in the surface arrangements and the surface showed a similar morphology to that existing in bulk.

  18. 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

  19. Structural strength of core graphite bars

    International Nuclear Information System (INIS)

    Kikuchi, K.; Futakawa, M.

    1987-01-01

    A HTR core consists of fuel, hot plenum, reflector and thermal barrier blocks. Each graphite block is supported by three thin cylindrical graphite bars called support post. Static and dynamic core loads are transmitted by the support posts to the thermal barrier blocks and a support plate. These posts are in contact with the blocks through hemispherical post seats to absorb the relative displacement caused by seismic force and the difference of thermal expansion of materials at the time of the start-up and shutdown of a reactor. The mixed fracture criterion of principal stress and modified Mohr-Coulomb's theory as well as the fracture criterion of principal stress based on elastic stress analysis was discussed in connection with the application to HTR graphite components. The buckling fracture of a support post was taken in consideration as one of the fracture modes. The effect that the length/diameter ratio of a post, small rotation and the curvature of post ends and seats exerted on the fracture strength was studied by using IG-110 graphite. Contacting stress analysis was carried out by using the structural analysis code 'COSMOS-7'. The experimental method, the analysis of buckling strength and the results are reported. The fracture of a support post is caused by the mixed mode of bending deformation, split fracture and shearing fracture. (Kako, I.)

  20. Characteristics of first loaded IG-110 graphite in HTTR core

    International Nuclear Information System (INIS)

    Sumita, Junya; Shibata, Taiju; Iyoku, Tatsuo; Sawa, Kazuhiro; Hanawa, Satoshi; Ishihara, Masahiro

    2006-10-01

    IG-110 graphite is a fine-grained isotropic and nuclear-grade graphite with excellent resistivity on both irradiation and corrosion and with high strength. The IG-110 graphite is used for the graphite components of High Temperature Engineering Test Reactor (HTTR) such as fuel and control rod guide blocks and support posts. In order to design and fabricate the graphite components in the HTTR, the Japan Atomic Energy Research Institute (the Japan Atomic Energy Agency at present) had established the graphite structural design code and design data on the basis of former research results. After the design code establishment, the IG-110 graphite components were fabricated and loaded in the HTTR core. This report summarized the characteristics of the first loaded IG-110 graphite as basic data for surveillance test, measuring material characteristics changed by neutron irradiation and oxidation. By comparing the design data, it was shown that the first loaded IG-110 graphite had excellent strength properties and enough safety margins to the stress limits in the design code. (author)

  1. Effects of Boron and Graphite Uncertainty in Fuel for TREAT Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Vaughn, Kyle; Mausolff, Zander; Gonzalez, Esteban; DeHart, Mark; Goluoglu, Sedat

    2017-03-01

    Advanced modeling techniques and current computational capacity make full core TREAT simulations possible, with the goal of such simulations to understand the pre-test core and minimize the number of required calibrations. But, in order to simulate TREAT with a high degree of precision the reactor materials and geometry must also be modeled with a high degree of precision. This paper examines how uncertainty in the reported values of boron and graphite have an effect on simulations of TREAT.

  2. Effect of graphite addition into mill scale waste as a potential bipolar plates material of proton exchange membrane fuel cells

    Science.gov (United States)

    Khaerudini, D. S.; Prakoso, G. B.; Insiyanda, D. R.; Widodo, H.; Destyorini, F.; Indayaningsih, N.

    2018-03-01

    Bipolar plates (BPP) is a vital component of proton exchange membrane fuel cells (PEMFC), which supplies fuel and oxidant to reactive sites, remove reaction products, collects produced current and provide mechanical support for the cells in the stack. This work concerns the utilization of mill scale, a by-product of iron and steel formed during the hot rolling of steel, as a potential material for use as BPP in PEMFC. On the other hand, mill scale is considered a very rich in iron source having characteristic required such as for current collector in BPP and would significantly contribute to lower the overall cost of PEMFC based fuel cell systems. In this study, the iron reach source of mill scale powder, after sieving of 150 mesh, was mechanically alloyed with the carbon source containing 5, 10, and 15 wt.% graphite using a shaker mill for 3 h. The mixed powders were then pressed at 300 MPa and sintered at 900 °C for 1 h under inert gas atmosphere. The structural changes of powder particles during mechanical alloying and after sintering were studied by X-ray diffractometry, optical microscopy, scanning electron microscopy, and microhardness measurement. The details of the presence of iron, carbon, and iron carbide (Fe-C) as the products of reactions as well as sufficient mechanical strength of the sintered materials were presented in this report.

  3. Effect of anode polarization on biofilm formation and electron transfer in Shewanella oneidensis/graphite felt microbial fuel cells.

    Science.gov (United States)

    Pinto, David; Coradin, Thibaud; Laberty-Robert, Christel

    2018-04-01

    In microbial fuel cells, electricity generation is assumed by bacterial degradation of low-grade organics generating electrons that are transferred to an electrode. The nature and efficiency of the electron transfer from the bacteria to the electrodes are determined by several chemical, physical and biological parameters. Specifically, the application of a specific potential at the bioanode has been shown to stimulate the formation of an electro-active biofilm, but the underlying mechanisms remain poorly understood. In this study, we have investigated the effect of an applied potential on the formation and electroactivity of biofilms established by Shewanella oneidensis bacteria on graphite felt electrodes in single- and double-chamber reactor configurations in oxic conditions. Using amperometry, cyclic voltammetry, and OCP/Power/Polarization curves techniques, we showed that a potential ranging between -0.3V and +0.5V (vs. Ag/AgCl/KCl sat.) and its converse application to a couple of electrodes leads to different electrochemical behaviors, anodic currents and biofilm architectures. For example, when the bacteria were confined in the anodic compartment of a double-chamber cell, a negative applied potential (-0.3V) at the bioanode favors a mediated electron transfer correlated with the progressive formation of a biofilm that fills the felt porosity and bridges the graphite fibers. In contrast, a positive applied potential (+0.3V) at the bioanode stimulates a direct electron transfer resulting in the fast-bacterial colonization of the fibers only. These results provide significant insight for the understanding of the complex bacteria-electrode interactions in microbial fuel cells. Copyright © 2017 Elsevier B.V. All rights reserved.

  4. Electrocatalytic properties of graphite nanofibers-supported platinum catalysts for direct methanol fuel cells.

    Science.gov (United States)

    Park, Soo-Jin; Park, Jeong-Min; Seo, Min-Kang

    2009-09-01

    Graphite nanofibers (GNFs) treated at various temperatures were used as carbon supports to improve the efficiency of PtRu catalysts. The electrochemical properties of the PtRu/GNFs catalysts were then investigated to evaluate their potential for application in DMFCs. The results indicated that the particle size and dispersibility of PtRu in the catalysts were changed by heat treatment, and the electrochemical activity of the catalysts was improved. Consequently, it was found that heat treatments could have an influence on the surface and structural properties of GNFs, resulting in enhancing an electrocatalytic activity of the catalysts for DMFCs.

  5. 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.

  6. Seismic response of high temperature gas-cooled reactor core with block-type fuel, (2)

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1980-01-01

    For the aseismic design of a high temperature gas-cooled reactor (HTGR) with block-type fuel, it is necessary to predict the motion and force of core columns and blocks. To reveal column vibration characteristics in three-dimensional space and impact response, column vibration tests were carried out with a scale model of a one-region section (seven columns) of the HTGR core. The results are as follows: (1) the column has a soft spring characteristic based on stacked blocks connected with loose pins, (2) the column has whirling phenomena, (3) the compression spring force simulating the gas pressure has the effect of raising the column resonance frequency, and (4) the vibration behavior of the stacked block column and impact response of the surrounding columns show agreement between experiment and analysis. (author)

  7. Effects of fuel relocation on reflood in a partially-blocked rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Jae [School of Mechanical Engineering, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 34134 (Korea, Republic of); Kim, Jongrok; Kim, Kihwan; Bae, Sung Won [Thermal-Hydraulic Safety Research Division, Korea Atomic Energy Research Division, 111 Daedeok-daero, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Moon, Sang-Ki, E-mail: skmoon@kaeri.re.kr [Thermal-Hydraulic Safety Research Division, Korea Atomic Energy Research Division, 111 Daedeok-daero, Yuseong-gu, Daejeon 34057 (Korea, Republic of)

    2017-02-15

    Ballooning of the fuel rods has been an important issue, since it can influence the coolability of the rod bundle in a large-break loss-of-coolant accident (LBLOCA). Numerous past studies have investigated the effect of blockage geometry on the heat transfer in a partially blocked rod bundle. However, they did not consider the occurrence of fuel relocation and the corresponding effect on two-phase heat transfer. Some fragmented fuel particles located above the ballooned region may drop into the enlarged volume of the balloon. Accordingly, the fuel relocation brings in a local power increase in the ballooned region. The present study’s objective is to investigate the effect of the fuel relocation on the reflood under a LBLOCA condition. Toward this end, experiments were performed in a 5 × 5 partially-blocked rod bundle. Two power profiles were tested: one is a typical cosine shape and the other is the modified shape considering the effect of the fuel relocation. For a typical power shape, the peak temperature in the ballooned rods was lower than that in the intact rods. On the other hand, for the modified power shape, the peak temperature in the ballooned rods was higher than that in the intact rods. Numerical simulations were also performed using the MARS code. The tendencies of the peak clad temperatures were well predicted.

  8. US/FRG umbrella agreement for cooperation in GCR Development. Fuel, fission products, and graphite subprogram. Quarterly status report, July 1, 1982-September 30, 1982

    International Nuclear Information System (INIS)

    Turner, R.F.

    1982-10-01

    This report describes the status of the cooperative work being performed in the Fuel, Fission Product, and Graphite Subprogram under the HTR-Implementing Agreement of the United States/Federal Republic of Germany Umbrella Agreement for Cooperation in GCR Development. The status is described relative to the commitments in the Subprogram Plan for Fuel, Fission Products, and Graphite, Revision 5, April 1982. The work described was performed during the period July 1, 1982 through September 30, 1982 in the HTGR Base Technology Program at Oak Ridge National Laboratory, the HTGR Fuel and Plant Technology Programs at General Atomic Company (GA), and the Project HTR-Brennstoffkreislauf of the Entwicklungsgemeinschaft HTR at KFA Julich, HRB Mannheim, HOBEG Hanau, and SIGRI Meitingen. The requirement for and format of this quarterly status report are specified in the HTR Implementing Agreement procedures for cooperation. Responsibility for preparation of the quarterly report alternates between GA and KFA

  9. Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    2002-01-01

    The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [∼1000 Mw(e)] with passive safety systems to provide the potential for improved economics

  10. CFD analysis of blockage length on a partially blocked fuel rod

    International Nuclear Information System (INIS)

    Scuro, Nikolas Lymberis; Andrade, Delvonei Alves de; Angelo, Gabriel; Angelo, Edvaldo

    2017-01-01

    In LOCA accidents, fuel rods may balloon by the increasing of pressure difference between fuel rod and core vessel. With the balloon effect, the swelling can partially block the flow channel, affecting the coolability during reflood phase. In order to analyze the influence of blockage length after LOCA events, many numerical simulations using Ansys-CFX code have been done in steady state condition, characterizing the final phase of reflood. Peaks of temperature are observed in the middle of the fuel rod, followed by a temperature drop. This effect is justified by the increasing of heat transfer coefficient, originated from the high turbulence effects. Therefore, this paper considers a radial blockage of 90%, varying just the blockage length. This study observed that, for the same boundary conditions, the longer the blockage length originated after LOCA events, the higher are the central temperatures in the fuel rod. (author)

  11. CFD analysis of blockage length on a partially blocked fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Scuro, Nikolas Lymberis; Andrade, Delvonei Alves de [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear; Angelo, Gabriel [Centro Universitário FEI (UNIFEI), São Paulo, SP (Brazil). Dept. de Engenharia Mecânica; Angelo, Edvaldo, E-mail: nikolas.scuro@gmail.com, E-mail: delvonei@ipen.br, E-mail: gangelo@fei.edu.br, E-mail: eangelo@mackenzie.br [Universidade Presbiteriana Mackenzie, São Paulo, SP (Brazil). Escola da Engenharia. Grupo de Simulação Numérica

    2017-07-01

    In LOCA accidents, fuel rods may balloon by the increasing of pressure difference between fuel rod and core vessel. With the balloon effect, the swelling can partially block the flow channel, affecting the coolability during reflood phase. In order to analyze the influence of blockage length after LOCA events, many numerical simulations using Ansys-CFX code have been done in steady state condition, characterizing the final phase of reflood. Peaks of temperature are observed in the middle of the fuel rod, followed by a temperature drop. This effect is justified by the increasing of heat transfer coefficient, originated from the high turbulence effects. Therefore, this paper considers a radial blockage of 90%, varying just the blockage length. This study observed that, for the same boundary conditions, the longer the blockage length originated after LOCA events, the higher are the central temperatures in the fuel rod. (author)

  12. Biodegradation of phenol in batch and continuous flow microbial fuel cells with rod and granular graphite electrodes.

    Science.gov (United States)

    Moreno, Lyman; Nemati, Mehdi; Predicala, Bernardo

    2018-01-01

    Phenol biodegradation was evaluated in batch and continuous flow microbial fuel cells (MFCs). In batch-operated MFCs, biodegradation of 100-1000 mg L -1 phenol was four to six times faster when graphite granules were used instead of rods (3.5-4.8 mg L -1  h -1 vs 0.5-0.9 mg L -1  h -1 ). Similarly maximum phenol biodegradation rates in continuous MFCs with granular and single-rod electrodes were 11.5 and 0.8 mg L -1  h -1 , respectively. This superior performance was also evident in terms of electrochemical outputs, whereby continuous flow MFCs with granular graphite electrodes achieved maximum current and power densities (3444.4 mA m -3 and 777.8 mW m -3 ) that were markedly higher than those with single-rod electrodes (37.3 mA m -3 and 0.8 mW m -3 ). Addition of neutral red enhanced the electrochemical outputs to 5714.3 mA m -3 and 1428.6 mW m -3 . Using the data generated in the continuous flow MFC, biokinetic parameters including μ m , K S , Y and K e were determined as 0.03 h -1 , 24.2 mg L -1 , 0.25 mg cell (mg phenol) -1 and 3.7 × 10 -4  h -1 , respectively. Access to detailed kinetic information generated in MFC environmental conditions is critical in the design, operation and control of large-scale treatment systems utilizing MFC technology.

  13. 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)

  14. CFD Analysis of Hot Spot Fuel Temperature in the Control Fuel Block Assembly of a VHTR core

    International Nuclear Information System (INIS)

    Kim, Min Hwan; Tak, Nam Il; Noh, Jae Man

    2010-01-01

    The Very High Temperature Reactor (VHTR) dedicated for efficient hydrogen production requires core outlet temperatures of more than 950 .deg. C. As the outlet temperature increases, the thermal margin of the core decreases, which highlights the need for a detailed analysis to reduce its uncertainty. Tak et al. performed CFD analysis for a 1/12 fuel assembly model and compared the result with a simple unit-cell model in order to emphasize the need of a detailed CFD analysis for the prediction of hot spot fuel temperatures. Their CFD model, however, was focused on the standard fuel assembly but not on the control fuel assembly in which a considerable amount of bypass flow is expected to occur through the control rod passages. In this study, a CFD model for the control fuel block assembly is developed and applied for the hot spot analyses of PMR200 core. Not only the bypass flow but also the cross flow is considered in the analyses

  15. Evaluation of Effective thermal conductivity models on the prismatic fuel block of a Very High Temperature Reactor by CFD analysis

    International Nuclear Information System (INIS)

    Shin, Dong-Ho; Cho, Hyoung-Kyu; Tak, Nam-Il; Park, Goon-Cherl

    2014-01-01

    Effective thermal conductivity models which can be used to analyze the heat transfer phenomena of a prismatic fuel block were evaluated by CFD analysis. In the accident condition of VHTR when forced convection is lost, the heat flows in radial direction through the hexagonal fuel blocks that contain the large number of coolant holes and fuel compacts. Due to the complex geometry of fuel block and radiation heat transfer; the detail heat transfer computation on the fuel block needs excessive computation resources. Therefore, the detail computation isn’t appropriate for the lumped parameter code. The system code such as GAMMA+ adopts effective thermal conductivity model. Despite the complexity in heat transfer modes, the accurate analysis on the heat transfer in fuel block is necessary since it is directly relevant to the integrity of nuclear fuel embedded in fuel block. To satisfy the accurate analysis of complex heat transfer modes with limited computing sources, the credible effective thermal conductivity (ETC) models in which the effects of all of heat transfer modes are lumped is necessary. In this study, various ETC models were introduced and they are evaluated with CFD calculations. It is estimated that Maxwell-based model was the most pertinent one among the introduced ETC models. (author)

  16. 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.

  17. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope

    2011-10-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  18. Direct methanol fuel cell with extended reaction zone anode: PtRu and PtRuMo supported on graphite felt

    Science.gov (United States)

    Bauer, Alex; Gyenge, Előd L.; Oloman, Colin W.

    Pressed graphite felt (thickness ∼350 μm) with electrodeposited PtRu (43 g m -2, 1.4:1 atomic ratio) or PtRuMo (52 g m -2, 1:1:0.3 atomic ratio) nanoparticle catalysts was investigated as an anode for direct methanol fuel cells. At temperatures above 333 K the fuel cell performance of the PtRuMo catalyst was superior compared to PtRu. The power density was 2200 W m -2 with PtRuMo at 5500 A m -2 and 353 K while under the same conditions PtRu yielded 1925 W m -2. However, the degradation rate of the Mo containing catalyst formulation was higher. Compared to conventional gas diffusion electrodes with comparable PtRu catalyst composition and load, the graphite felt anodes gave higher power densities mainly due to the extended reaction zone for methanol oxidation.

  19. Diffusivities of Ag, Cs, Sr, and Kr in TRISO fuel particles and graphite

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Tri-structural isotropic (TRISO) coated particles have been developed and studied since the late 1950s when the concept of coated particles was invented by Roy Huddle of the United Kingdom Atomic Energy Authority. Several decades of work by half a dozen countries on fission product transport in TRISO fuel through numerous irradiation and heating experiments have led to several recommendations of transport data and to the adoption of various sets of diffusion coefficients. In 1997, the International Atomic Energy Agency (IAEA) gathered all these historical results and issued a technical document (TECDOC-978 [IAEA]) that summarizes these sets of recommended diffusion coefficients. Table 1 shows the reference literature articles for the diffusivities that have historically been recommended by the American and German TRISO fuel development programs and that are summarized in the IAEA report (see section 7 for full references of these articles).

  20. Graphite coated with manganese oxide/multiwall carbon nanotubes composites as anodes in marine benthic microbial fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Fu, Yubin, E-mail: ffyybb@ouc.edu.cn; Yu, Jian; Zhang, Yelong; Meng, Yao

    2014-10-30

    Highlights: • MnO{sub 2}/MWCNTs composites anode exhibits faster reaction kinetics. • The surfaces of MnO{sub 2}/MWCNTs composites anode exhibits better wettability. • A BMFC using the modified anode have excellent power output. - Abstract: Improving anode performance is of great significance to scale up benthic microbial fuel cells (BMFCs) for its marine application to drive oceanography instruments. In this study, manganese oxide (MnO{sub 2})/multiwall carbon nanotubes (MWCNTs) composites are prepared to be as novel anodes in the BMFCs via a direct redox reaction between permanganate ions (MnO{sub 4}{sup −}) and MWCNTs. The results indicate that the MnO{sub 2}/MWCNTs anode has a better wettability, greater kinetic activity and higher power density than that of the plain graphite (PG) anode. It is noted that the MnO{sub 2} (50% weight percent)/MWCNTs anode shows the highest electrochemical performance among them and will be a promising material for improving bioelectricity production of the BMFCs. Finally, a synergistic mechanism of electron transfer shuttle of Mn ions and their redox reactions in the interface between modified anode and bacteria biofilm are proposed to explain its excellent electrochemical performance.

  1. Performances of helium, neon and argon glow discharges for reduction of fuel hydrogen retention in tungsten, stainless steel and graphite

    International Nuclear Information System (INIS)

    Hino, T.; Yamauchi, Y.; Kimura, Y.; Matsumoto, A.; Nishimura, K.; Ueda, Y.

    2012-11-01

    It is quite important to investigate the performance of glow discharge conditionings for controls of in-vessel tritium (T) inventory and hydrogen recycling. For this purpose, first, the deuterium (D) retentions in tungsten (W), graphite (C) and stainless steel (SS) were measured. The retention in W was not small as expected, several times larger than that of SS, although the retention in SS was one order smaller than that of C. Such the large retention in W is owing to the growth of rough surface structure produced by plasma irradiations. For reduction of deuterium retention in W, SS and C, second, inert gas (He, Ne, Ar) glow discharges were conducted under the same condition, and these performances were compared. The removal ratio of deuterium retention was highest in He discharge, and lowest in Ar discharge. These values are well explained by the numerical analyses using SRIM code. The removal ratios for SS and C were significantly large, but quite small for W. This reason is again owing to the rough surface structure in W. For W, thirdly, the hydrogen isotope exchange and the wall baking experiments were conducted. It is found that the wall backing with a temperature higher than 700 K can well reduce the retention, and the hydrogen isotope exchange using deuterium glow discharge is also useful to reduce the tritium retention in the wall. The present results significantly contribute to control the fuel hydrogen retention and to reduce the in-vessel tritium inventory in fusion reactors. (author)

  2. High performance nano-Ni/Graphite electrode for electro-oxidation in direct alkaline ethanol fuel cells

    Science.gov (United States)

    Soliman, Ahmed B.; Abdel-Samad, Hesham S.; Abdel Rehim, Sayed S.; Ahmed, Mohamed A.; Hassan, Hamdy H.

    2016-09-01

    Ni/Graphite electrocatalysts (Ni/G) are successfully prepared through electrodeposition of Ni from acidic (pH = 0.8) and feebly acidic (pH = 5.5) aqueous Ni (II) baths. The efficiencies of such electrodes are investigated as anodes for direct alkaline ethanol fuel cells through their ethanol electrooxidation cyclic voltammetric (CV) response in alkaline medium. A direct proportionality between the amount of the electrodeposited Ni and its CV response is found. The amounts of the deposited Ni from the two baths are recorded using the Electrochemical Quartz Crystal Microbalance (eQCM). The Ni/G electrodes prepared from the feebly acidic bath show a higher electrocatalytic response than those prepared from the acidic bath. Surface morphology of the Ni particles electrodeposited from feebly acidic bath appears in a nano-scale dimension. Various electrochemical experiments are conducted to confirm that the Ni/G ethanol electrooxidation CV response greatly depends on the pH rather than nickel ion concentration of the deposition bath. The eQCM technique is used to detect the crystalline phases of nickel as α-Ni(OH)2/γ-NiOOH and β-Ni(OH)2/β-NiOOH and their in-situ inter-transformations during the potentiodynamic polarization.

  3. Macroporous graphitic carbon foam decorated with polydopamine as a high-performance anode for microbial fuel cell

    Science.gov (United States)

    Jiang, Hongmei; Yang, Lu; Deng, Wenfang; Tan, Yueming; Xie, Qingji

    2017-09-01

    Herein, a macroporous graphitic carbon foam (MGCF) electrode decorated with polydopamine (PDA) is used as a high-performance anode for microbial fuel cell (MFC) applications. The MGCF is facilely prepared by pyrolysis of a powder mixture comprising maltose, nickel nitrate, and ammonia chloride, without using solid porous template. The MGCF is coated with PDA by self-polymerization of dopamine in a basic solution. The MGCF can provide a large surface area for bacterial attachment, and PDA coated on the MGCF electrode can further promote bacterial adhesion resulting from the improved hydrophility, so the MGCF-PDA electrode as an anode in a MFC can show ultrahigh bacterial loading capacity. Moreover, the electrochemical oxidation of flavins at the MGCF-PDA electrode is greatly accelerated, so the extracellular electron transfer mediated by flavins is improved. As a result, the MFC equipped with a MGCF-PDA anode can show a maximum power density of 1735 mW cm-2, which is 6.7 times that of a MFC equipped with a commercial carbon felt anode, indicating a promising anode for MFC applications.

  4. Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

    2014-12-15

    Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

  5. 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

  6. 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

  7. Equilibrium core layout for the 1000 MW direct cycle HTR (HHT) with hexagonal monolith moulded fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Dworak, A

    1973-03-15

    The aim of this survey is to calculate an equilibrium Thorium fuel cycle for a 1000 MW HHT-core in off-load refuelling with hexagonal monolith moulded fuel blocks. It was tried to achieve an axial power distribution similar to the advanced pebble-bed reactors (OTTO) by introducing three axial core zones with different heavy metal content and initial enrichment.

  8. Artificial fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hamon, L L.W.

    1918-08-20

    Lignite, peat, sud, leaf-mold, or shale, or two or more of these raw carbonaceous materials are mixed with cellulose material, such as sawdust, silica, alkali, and tar or pitch, or residues from tar or pitch, or residues from the distillation of oils, and the mixture is molded into blocks. Other carbonaceous materials, such as graphite, anthracite, or coal-dust, coke, breeze, or culm, and mineral substances, such as iron and manganese ores, may be added. A smokeless fuel can be obtained by coking the blocks in the usual way in retorts.

  9. 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.

  10. Gas-Cooled Thorium Reactor with Fuel Block of the Unified Design

    Directory of Open Access Journals (Sweden)

    Igor Shamanin

    2015-01-01

    Full Text Available Scientific researches of new technological platform realization carried out in Russia are based on ideas of nuclear fuel breeding in closed fuel cycle and physical principles of fast neutron reactors. Innovative projects of low-power reactor systems correspond to the new technological platform. High-temperature gas-cooled thorium reactors with good transportability properties, small installation time, and operation without overloading for a long time are considered perspective. Such small modular reactor systems at good commercial, competitive level are capable of creating the basis of the regional power industry of the Russian Federation. The analysis of information about application of thorium as fuel in reactor systems and its perspective use is presented in the work. The results of the first stage of neutron-physical researches of a 3D model of the high-temperature gas-cooled thorium reactor based on the fuel block of the unified design are given. The calculation 3D model for the program code of MCU-5 series was developed. According to the comparison results of neutron-physical characteristics, several optimum reactor core compositions were chosen. The results of calculations of the reactivity margins, neutron flux distribution, and power density in the reactor core for the chosen core compositions are presented in the work.

  11. Silver/iron oxide/graphitic carbon composites as bacteriostatic catalysts for enhancing oxygen reduction in microbial fuel cells

    Science.gov (United States)

    Ma, Ming; You, Shijie; Gong, Xiaobo; Dai, Ying; Zou, Jinlong; Fu, Honggang

    2015-06-01

    Biofilms from anode heterotrophic bacteria are inevitably formed over cathodic catalytic sites, limiting the performances of single-chamber microbial fuel cells (MFCs). Graphitic carbon (GC) - based nano silver/iron oxide (AgNPs/Fe3O4/GC) composites are prepared from waste pomelo skin and used as antibacterial oxygen reduction catalysts for MFCs. AgNPs and Fe3O4 are introduced in situ into the composites by one-step carbothermal reduction, enhancing their conductivity and catalytic activity. To investigate the effects of Fe species on the antibacterial and catalytic properties, AgNPs/Fe3O4/GC is washed with sulfuric acid (1 mol L-1) for 0.5 h, 1 h, and 5 h and marked as AgNPs/Fe3O4/GC-x (x = 0.5 h, 1 h and 5 h, respectively). A maximum power density of 1712 ± 35 mW m-2 is obtained by AgNPs/Fe3O4/GC-1 h, which declines by 4.12% after 17 cycles. Under catalysis of all AgNP-containing catalysts, oxygen reduction reaction (ORR) proceeds via the 4e- pathway, and no toxic effects to anode microorganisms result from inhibiting the cathodic biofilm overgrowth. With the exception of AgNPs/Fe3O4/GC-5 h, the AgNPs-containing composites exhibit remarkable power output and coulombic efficiency through lowering proton transfer resistance and air-cathode biofouling. This study provides a perspective for the practical application of MFCs using these efficient antibacterial ORR catalysts.

  12. An enhanced model for minimizing fuel consumption under block-queuing in a drive-through service system

    Energy Technology Data Exchange (ETDEWEB)

    Reilly, C.H.; Berglin, J. [University of Central Florida, Orlando, FL (United States). Dept. of Industrial Engineering and Management Systems

    2004-05-01

    We present a new model for determining the optimal block-size under block-queuing in a simple, single-channel queue at a drive-through service facility. With block-queuing, a queue is partitioned into an active section and a passive section, where drivers are asked to turn off their engines until the active section clears. Our model prescribes a block-size, i.e., a maximum number of vehicles in the active section, which minimizes the expected amount of fuel consumed in the queue. It can assess the effects of the traffic intensity, the service-time variance, and the proportion of compliant drivers in the passive section on the optimal block- size and on fuel consumption in the queue. (author)

  13. 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

  14. 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

  15. Development of a dry-mechanical graphite separation process and elimination of the separated carbon for the reprocessing of spherical HTR fuel elements

    International Nuclear Information System (INIS)

    Kronschnabel, H.

    1982-01-01

    Due to the C-14 distribution the separation of the particle-free outer region of the spherical HTR fuel element with subsequent solidification of the separated carbon makes it possible to reduce by half the remaining C-14 inventory in the inner particle region to be further treated. Separation of the particle-free outer region by a newly developed sphere-peeling milling machine, conditioning the graphite into compacts and in-situ cementation into a salt-mine are the basic elements of this head-end process variation. An annual cavern volume of approx. 2000 m 3 will be needed to ultimately store the graphite of the particle-free outer region, which corresponds to a reprocessing capacity of 50 GWsub(e) installed HTR power. The brush-disintegration of the remaining inner particle region and the resulting peel-brush-preparation are capable of separating 95% of the graphite without any heavy metal losses. With the mentioned reprocessing capacity an annual cavern volume of approx. 16.500 m 3 is required. (orig.) [de

  16. Optimization of axial enrichment distribution for BWR fuels using scoping libraries and block coordinate descent method

    Energy Technology Data Exchange (ETDEWEB)

    Tung, Wu-Hsiung, E-mail: wstong@iner.gov.tw; Lee, Tien-Tso; Kuo, Weng-Sheng; Yaur, Shung-Jung

    2017-03-15

    Highlights: • An optimization method for axial enrichment distribution in a BWR fuel was developed. • Block coordinate descent method is employed to search for optimal solution. • Scoping libraries are used to reduce computational effort. • Optimization search space consists of enrichment difference parameters. • Capability of the method to find optimal solution is demonstrated. - Abstract: An optimization method has been developed to search for the optimal axial enrichment distribution in a fuel assembly for a boiling water reactor core. The optimization method features: (1) employing the block coordinate descent method to find the optimal solution in the space of enrichment difference parameters, (2) using scoping libraries to reduce the amount of CASMO-4 calculation, and (3) integrating a core critical constraint into the objective function that is used to quantify the quality of an axial enrichment design. The objective function consists of the weighted sum of core parameters such as shutdown margin and critical power ratio. The core parameters are evaluated by using SIMULATE-3, and the cross section data required for the SIMULATE-3 calculation are generated by using CASMO-4 and scoping libraries. The application of the method to a 4-segment fuel design (with the highest allowable segment enrichment relaxed to 5%) demonstrated that the method can obtain an axial enrichment design with improved thermal limit ratios and objective function value while satisfying the core design constraints and core critical requirement through the use of an objective function. The use of scoping libraries effectively reduced the number of CASMO-4 calculation, from 85 to 24, in the 4-segment optimization case. An exhausted search was performed to examine the capability of the method in finding the optimal solution for a 4-segment fuel design. The results show that the method found a solution very close to the optimum obtained by the exhausted search. The number of

  17. Lurgi MegaMethanol technology. Delivering the building blocks for the future fuel and monomer demand

    Energy Technology Data Exchange (ETDEWEB)

    Wurzel, T. [Lurgi AG, Frankfurt/Main (Germany)

    2006-07-01

    The paper describes the central role of methanol within a changing environment with respect to feedstock availability as well as steadily growing demand in fuel and monomer demand. The current large-scale production facilities are described with respect to the technological challenges in order to ensure the availability of sufficient methanol for down-stream applications. Different down-stream applications are described which clearly confirm that methanol is the dominant C1-building block due to its chemical flexibility. It is concluded that by means of the implementation of two MTP (Methanol to Propylene) projects in China initiated the era of ''down-stream methanol'' has begun in the industry. (orig.)

  18. French programme for HTR fuel

    International Nuclear Information System (INIS)

    Gillet, R.M.

    1991-01-01

    It is reported that in the frameworks of the French HTR research program, stopped in 1979 the HTR coated particle fuel, fuel rod and prismatic fuel element design have been successfully developed and irradiation tested in France and specific examination methods for irradiated fuel particles, rods and graphite blocks have been developed. Currently CEA is involved in fission product transport experiments sponsored by the US Department of Energy and performed in the COMEDIE loop. Finally the CEA follows progress and developments in HTR fuel research and development throughout the world. 1 tab

  19. Functioning of blocks of compacted bentonite in a repository for spent nuclear fuels

    International Nuclear Information System (INIS)

    Boergesson, Lennart; Kalbantner, P.; Sjoeblom, R.

    2001-12-01

    The main purpose of the presented work is to provide The Swedish Nuclear Fuel and Waste Management Company (SKB) with a proposed set of requirements regarding the functioning of the blocks of compacted bentonite. These blocks are intended to constitute the bentonite envelope which after uptake of water will form the buffer between the canister and the rock. The purpose is also to provide a basis for SKB for their direction of the continued development work for the selection of a reference technology and for creating a quality system for the buffer material. No attempts are made in the report to derive the functional requirements. Instead, such requirements are postulated based on realistic scenarios regarding the chain of processes from excavation - transport - preparation of press powder - compaction - handling and emplacement in the deposition hole. It is the strategy of SKB to use a natural material which after the above-mentioned processes forms a buffer with properties which closely resemble those of the original material. This implies that all process steps must be designed in such a way that the properties of the bentonite do not change to any significant degree with respect to the disposal function. The main results in the report are as follows: A set of functional requirements are compiled and presented. These concord with the different descriptions given on the process steps. The requirements are generic and are assessed to be relatively invariant for various operational requirements and process controls. The process chain comprising excavation of bentonite - transport - preparation of press powder - compaction - handling and emplacement are explained. The presentations of functional requirements and processes are foreseen to constitute a basis for a comparison between uniaxial and isostatic compaction and can be an important basis for SKB's quality work. The development of cracks in the bentonite blocks has been identified as an important aspect for the

  20. Direct methanol fuel cell with extended reaction zone anode: PtRu and PtRuMo supported on graphite felt

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, Alex; Gyenge, Elod L.; Oloman, Colin W. [Department of Chemical and Biological Engineering, The University of British Columbia, 2360 East Mall, Vancouver, BC (Canada)

    2007-05-15

    Pressed graphite felt (thickness {proportional_to}350 {mu}m) with electrodeposited PtRu (43 g m{sup -2}, 1.4:1 atomic ratio) or PtRuMo (52 g m{sup -2}, 1:1:0.3 atomic ratio) nanoparticle catalysts was investigated as an anode for direct methanol fuel cells. At temperatures above 333 K the fuel cell performance of the PtRuMo catalyst was superior compared to PtRu. The power density was 2200 W m{sup -2} with PtRuMo at 5500 A m{sup -2} and 353 K while under the same conditions PtRu yielded 1925 W m{sup -2}. However, the degradation rate of the Mo containing catalyst formulation was higher. Compared to conventional gas diffusion electrodes with comparable PtRu catalyst composition and load, the graphite felt anodes gave higher power densities mainly due to the extended reaction zone for methanol oxidation. (author)

  1. Preparation of anode-electrolyte structures using graphite, sodium bicarbonate or citric acid as pore forming agents for application in solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Paz Fiuza, Raigenis da; Silva, Marcos Aurelio da; Guedes, Bruna C.; Pontes, Luiz A.; Boaventura, Jaime Soares [UFBA, Salvador, Bahia (Brazil). Energy and Materials Science Group

    2010-07-01

    Cermets based on Ni supported on YSZ or GDC were prepared for use as anode in direct reform SOFCs. NaHCO3 (Na-Ni-YSZ and Na-Ni-GDC) or citric acid (Ac-Ni-YSZ and Ac-Ni-GDC) were used as pore forming agents (PFAs). The SOFC anode was also prepared using graphite (G-Ni-YSZ and G-Ni-GDC) as PFA for the purposes of comparison. The testing unitary SOFC, planar type, was made by pressing the anode-electrolyte assembly, followed by sintering at 1500 C. After this, LSM (lanthanum and strontium manganite) paint was used for the cathode deposition. The powdered cermets were evaluated in ethanol steam reforming at 650 C. The ethanol conversion was 84% and 32% for cermets Na-Ni-YSZ and G-Ni-YSZ, respectively and the selectivity to H{sub 2} was 32 and 20% for the two cermets, respectively. The Na-Ni-YSZ cermet was ten times more resistant to carbon deposition than the G-Ni-YSZ cermet. SEM micrographs of the anode-electrolyte assembly showed that the use of NaHCO{sub 3} as PFA created a well formed interface between layers with homogeneously distributed pores. In contrast, graphite as PFA formed a loose interface between anode and electrolyte. The performance of the unitary SOFC was evaluated using ethanol, hydrogen or methane as fuel. The cell operated well using any of these fuels; however, they exhibited different electrochemical behavior. (orig.)

  2. Measurement of the enthalpy and specific heat of a Be2C-graphite-UC2 reactor fuel material to 19800K

    International Nuclear Information System (INIS)

    Roth, E.P.

    1980-01-01

    The enthalpy and specific heat of a Be 2 C-graphite-UC 2 composite nuclear fuel material were measured over the temperature range 300 to 1980 0 K using differential scanning calorimetry and liquid argon vaporization calorimetry. The fuel material measured was developed at Sandia National Laboratories for use in pulsed test reactors. The material is a hot-pressed composite consisting of 40 vol % Be 2 C, 49.5 vol % graphite, 3.5 vol % UC 2 and 7.0 vol % void. The specific heat was measured with the differential scanning calorimeter over the temperature range 300 to 950 0 K while the enthalpy was measured over the range 1185 to 1980 0 K with the liquid argon vaporization calorimeter. The normal spectral emittance at a wavelength of 6.5 x 10 -5 cm was measured over the experimental temperature range. The combined experimental enthalpy data were fit using a spline routine and differentiated to give the specific heat. Comparison of the measured specific heat of the composite to the specific heat calculated by summing the contributions of the individual components indicates that the specific heat of the Be 2 C component differs significantly from literature values and is approximately 0.6 cal/g-K (2.5 x 10 3 J/Kg-K) for temperatures above 1000 0 K

  3. On the Deposition Equilibrium of Carbon Nanotubes or Graphite in the Reforming Processes of Lower Hydrocarbon Fuels

    Directory of Open Access Journals (Sweden)

    Zdzisław Jaworski

    2017-11-01

    Full Text Available The modeling of carbon deposition from C-H-O reformates has usually employed thermodynamic data for graphite, but has rarely employed such data for impure filamentous carbon. Therefore, electrochemical data for the literature on the chemical potential of two types of purified carbon nanotubes (CNTs are included in the study. Parameter values determining the thermodynamic equilibrium of the deposition of either graphite or CNTs are computed for dry and wet reformates from natural gas and liquefied petroleum gas. The calculation results are presented as the atomic oxygen-to-carbon ratio (O/C against temperature (200 to 100 °C for various pressures (1 to 30 bar. Areas of O/C for either carbon deposition or deposition-free are computed, and indicate the critical O/C values below which the deposition can occur. Only three types of deposited carbon were found in the studied equilibrium conditions: Graphite, multi-walled CNTs, and single-walled CNTs in bundles. The temperature regions of the appearance of the thermodynamically stable forms of solid carbon are numerically determined as being independent of pressure and the analyzed reactants. The modeling indicates a significant increase in the critical O/C for the deposition of CNTs against that for graphite. The highest rise in the critical O/C, of up to 290% at 30 bar, was found for the wet reforming process.

  4. Electrochemically assisted organosol method for Pt-Sn nanoparticle synthesis and in situ deposition on graphite felt support: Extended reaction zone anodes for direct ethanol fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Lycke, Derek R.; Gyenge, Elod L. [Department of Chemical and Biological Engineering, The University of British Columbia, 2360 East Mall, Vancouver, BC (Canada)

    2007-03-20

    Two electrochemically assisted variants of the Boenneman organosol method were developed for Pt-Sn nanoparticle synthesis and in situ deposition on graphite felt electrodes (e.g. thickness up to 2 mm). Tetraoctylammonium triethylhydroborate N(C{sub 8}H{sub 17}){sub 4}BH(C{sub 2}H{sub 5}){sub 3} was employed as colloid stabilizer and reductant dissolved in tetrahydrofuran (THF). The role of the electric field at a low deposition current density of 1.25 mA cm{sup -2} was mainly electrophoretic causing the migration and adsorption of N(C{sub 8}H{sub 17}){sub 4}BH(C{sub 2}H{sub 5}){sub 3} on the graphite felt surface where it reduced the PtCl{sub 2}-SnCl{sub 2} mixture. Faradaic electrodeposition was detected mostly for Sn. Typical Pt-Sn loadings were between 0.4 and 0.9 mg cm{sup -2} depending on the type of pre-deposition exposure of the graphite felt: surfactant-adsorption and metal-adsorption variant, respectively. The catalyst surface area and Pt:Sn surface area ratio was determined by anodic striping of an underpotential deposited Cu monolayer. The two deposition variants gave different catalyst surfaces: total area 233 and 76 cm{sup 2} mg{sup -1}, with Pt:Sn surface area ratio of 3.5:1 and 7.7:1 for surfactant and metal adsorption, respectively. Regarding electrocatalysis of ethanol oxidation, voltammetry and chronopotentiometry studies corroborated by direct ethanol fuel cell experiments using 0.5 M H{sub 2}SO{sub 4} as electrolyte, showed that due to a combination of higher catalyst load and Pt:Sn surface ratio, the graphite felt anodes prepared by the metal-adsorption variant gave better performance. The catalyzed graphite felt provided an extended reaction zone for ethanol electrooxidation and it gave higher catalyst mass specific peak power outputs compared to literature data obtained using gas diffusion anodes with carbon black supported Pt-Sn nanoparticles. (author)

  5. Graphitized nanodiamond supporting PtNi alloy as stable anodic and cathodic electrocatalysts for direct methanol fuel cell

    International Nuclear Information System (INIS)

    Wang, Yongjiao; Zang, Jianbing; Dong, Liang; Pan, Hong; Yuan, Yungang; Wang, Yanhui

    2013-01-01

    Highlights: • The graphitized nanodiamond (GND) showed a higher oxidation-resistance than XC-72. • The PtNi/GND electrocatalytic exhibited greater stability than PtNi/XC-72. • The PtNi/GND had a better catalytic activity for MOR and ORR than Pt/GND. -- Abstract: Surface graphitized nanodiamond (GND) with a diamond core covered by a graphitic carbon shell was prepared by annealing ND at the temperature of 1300 °C in a vacuum of 10 −3 Pa. PtNi electrocatalysts were prepared by a microwave heating polyol method using the prepared GND as a support. The composition and morphology of the PtNi electrocatalysts supported on GND (PtNi/GND) were characterized by X-ray diffraction, transmission electron microscopy and energy dispersion spectra. The results showed that nano-scaled PtNi alloy particles with an atomic ratio of approximately 1:1 were uniformly deposited on the GND through co-reduction process. The electrocatalytic activities of the PtNi/GND electrocatalysts for methanol oxidation reaction (MOR) and oxygen reduction reaction (ORR) were investigated by cyclic voltammetry, chronoamperometry and linear sweep voltammetry. The PtNi/GND exhibited better electrocatalytic activities than the Pt/GND either for MOR and ORR. In comparison with traditional carbon support Vulcan XC-72, GND showed higher oxidation-resistance, and consequently led to greater stability for the PtNi/GND than PtNi/XC-72

  6. 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

  7. Determination of molybdenum, ruthenium, rhodium, and palladium in radioinactive simulated waste of the nuclear fuel cycle by solid sampling graphite furnace atomic absorption spectrometry (GFAAS)

    International Nuclear Information System (INIS)

    Schmiedel, G.; Mainka, E.; Ache, H.J.

    1989-01-01

    In relation with insoluble particles in the nuclear fuel cycle waste, the solid sampling GFAAS was used to determine molybdenum, ruthenium, rhodium, and palladium in such waste. Two methods for the direct determination of these elements are described. The samples must be handled in glove boxes or moreover in hot cells with a robot. The determination of the elements by the cup-in-tube technique needs a very sensitive balance (microbalance) for weighing in μg-range and the handling of this method is not practical in glove boxes and hot cells. An alternative technique of solid sampling GFAAS, which can be used without great problems in glove boxes and hot cells is the slurry technique. In this case two methods have been used. One method uses graphite powder as a diluter, the other is the direct suspension of the sample in a matrix modifier solution. In the case of slurry technique with predilution of the sample with graphite powder, recoveries between 91 and 102% and RSD between 4 and 8% were obtained, whereas in the case of slurry technique with direct suspension of the waste sample recoveries between 91 and 103% and RSD between 14 and 20% for the above mentioned elements were obtained. (orig.)

  8. 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.

  9. Technical and regulatory review of the Rover nuclear fuel process for use on Fort St. Vrain fuel

    International Nuclear Information System (INIS)

    Hertzler, T.

    1993-02-01

    This report describes the results of an analysis for processing and final disposal of Fort St. Vrain (FSV) irradiated fuel in Rover-type equipment or technologies. This analysis includes an evaluation of the current Rover equipment status and the applicability of this technology in processing FSV fuel. The analyses are based on the physical characteristics of the FSV fuel and processing capabilities of the Rover equipment. Alternate FSV fuel disposal options are also considered including fuel-rod removal from the block, disposal of the empty block, or disposal of the entire fuel-containing block. The results of these analyses document that the current Rover hardware is not operable for any purpose, and any effort to restart this hardware will require extensive modifications and re-evaluation. However, various aspects of the Rover technology, such as the successful fluid-bed burner design, can be applied with modification to FSV fuel processing. The current regulatory climate and technical knowledge are not adequately defined to allow a complete analysis and conclusion with respect to the disposal of intact fuel blocks with or without the fuel rods removed. The primary unknowns include the various aspects of fuel-rod removal from the block, concentration of radionuclides remaining in the graphite block after rod removal, and acceptability of carbon in the form of graphite in a high level waste repository

  10. 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

  11. Detection block

    International Nuclear Information System (INIS)

    Bezak, A.

    1987-01-01

    A diagram is given of a detection block used for monitoring burnup of nuclear reactor fuel. A shielding block is an important part of the detection block. It stabilizes the fuel assembly in the fixing hole in front of a collimator where a suitable gamma beam is defined for gamma spectrometry determination of fuel burnup. The detector case and a neutron source case are placed on opposite sides of the fixing hole. For neutron measurement for which the water in the tank is used as a moderator, the neutron detector-fuel assembly configuration is selected such that neutrons from spontaneous fission and neutrons induced with the neutron source can both be measured. The patented design of the detection block permits longitudinal travel and rotation of the fuel assembly to any position, and thus more reliable determination of nuclear fuel burnup. (E.S.). 1 fig

  12. Effect of Crossflow on Hot Spot Fuel Temperature in Prismatic VHTR

    International Nuclear Information System (INIS)

    Lee, Sung Nam; Tak, Nam-il; Kim, Min Hwan; Noh, Jae Man; Park, Goon-Cherl

    2014-01-01

    Various studies have been conducted to predict the thermal-hydraulics of a prismatic gas-cooled reactor. However, most previous studies have concentrated on the nominal-designed core. The fuel assembly of a high temperature gas-cooled reactor consists of a fuel compact and graphite block used as a moderator. This graphite faces a dimensional change due to irradiation or heating during normal operation. This size change might affect the coolant flow distribution in the active core. Therefore, the hot spot fuel temperature position or value could vary. There are two types of flows by the size change of graphite. One is the bypass flow and the other is the crossflow. The crossflow occurs at the crossflow gap between the vertical stacks of fuel blocks. In this study, the effect of the crossflow on the hot spot fuel temperature has been intensively investigated. (author)

  13. Neutron multipilication factors as a function of temperature: a comparison of calculated and measured values for lattices using 233UO2-ThO2 fuel in graphite

    International Nuclear Information System (INIS)

    Newman, D.F.; Gore, B.F.

    1978-01-01

    Neutron multiplication factors calculated as a function of temperature for three graphite-moderated 233 UO 2 -ThO 2 -fueled lattices are correlated with the values measured for these lattices in the high-temperature lattice test reactor (HTLTR). The correlation analysis is accomplished by fitting calculated values of k/sub infinity/(T) to the measured values using two least-squares-fitted correlation coefficients: (a) a normalization factor and (b) a temperature coefficient bias factor. These correlations indicate the existence of a negative (nonconservative) bias in temperature coefficients of reactivity calculated using ENDF/B-IV cross-section data. Use of an alternate cross-section data set for thorium, which has a smaller resonance integral than ENDF/B-IV data, improved the agreement between calculated and measured temperature coefficients of reactivity for the three experimental lattices. The results of the correlations are used to estimate the bias in the temperature coefficient of reactivity calculated for a lattice typical of fresh 233 U recycle fuel for a high-temperature gas-cooled reactor (HTGR). This extrapolation to a lattice having a heavier fissile loading than the experimental lattices is accomplished using a sensitivity analysis of the estimated bias to alternate thorium cross-section data used in calculations of k/sub infinity/(T). The envelope of uncertainty expected to contain the actual values for the temperature coefficient of the reactivity for the 233 U-fueled HTGR lattice studied remains negative at 1600 K (1327 0 C). Although a broader base of experimental data with improved accuracy is always desirable, the existing data base provided by the HTLTR experiments is judged to be adequate for the verification of neutronic calculations for the HTGR containing 233 U fuel at its current state of development

  14. Investigation of chemical modifiers for sulfur determination in diesel fuel samples by high-resolution continuum source graphite furnace molecular absorption spectrometry using direct analysis

    Energy Technology Data Exchange (ETDEWEB)

    Huber, Charles S. [Instituto Federal Sul-rio-grandense, Câmpus Pelotas, Pelotas, RS (Brazil); Universidade Federal do Rio Grande do Sul, Instituto de Química, Porto Alegre, RS (Brazil); Vale, Maria Goreti R. [Universidade Federal do Rio Grande do Sul, Instituto de Química, Porto Alegre, RS (Brazil); Instituto Nacional de Ciência e Tecnologia do CNPq, INCT de Energia e Ambiente, Universidade Federal da Bahia, Salvador, BA (Brazil); Welz, Bernhard [Instituto Nacional de Ciência e Tecnologia do CNPq, INCT de Energia e Ambiente, Universidade Federal da Bahia, Salvador, BA (Brazil); Universidade Federal de Santa Catarina, Departamento de Química, Florianópolis, SC (Brazil); Andrade, Jailson B. [Instituto Nacional de Ciência e Tecnologia do CNPq, INCT de Energia e Ambiente, Universidade Federal da Bahia, Salvador, BA (Brazil); Dessuy, Morgana B., E-mail: mbdessuy@ufrgs.br [Universidade Federal do Rio Grande do Sul, Instituto de Química, Porto Alegre, RS (Brazil); Instituto Nacional de Ciência e Tecnologia do CNPq, INCT de Energia e Ambiente, Universidade Federal da Bahia, Salvador, BA (Brazil)

    2015-06-01

    High-resolution continuum source graphite furnace molecular absorption spectrometry has been applied for sulfur determination in diesel fuel. The sharp rotational lines of the carbon monosulfide molecule (formed during the vaporization step) were used to measure the absorbance. The analytical line at 258.056 nm was monitored using the sum of three pixels. Different chemical modifiers were investigated and the mixture of palladium and magnesium was used as chemical modifier in combination with iridium as permanent modifier. L-Cysteine was chosen as sulfur standard and the calibration was done against aqueous standard solutions. The proposed method was applied for the analyses of four diesel samples: two S10 samples and two S500 samples. The trueness of the method was checked with a certified reference material (CRM) of sulfur in diesel fuel (NIST 2724b). Accurate results, for samples and CRM, were achieved after a dilution with propan-1-ol. The following figures of merit were obtained: characteristic mass of 17 ± 3 ng, limit of detection and limit of quantification of 1.4 mg kg{sup −1} and 4.7 mg kg{sup −1}, respectively. - Highlights: • Ir, Ru and Zr were investigated as permanent modifiers. • Ca, Mg, Pd and Pd/Mg were investigated as modifiers in solution. • Indirect determination of sulfur monitoring the molecular absorbance of the CS • Direct analysis of diesel samples using a dilution in propan-1-ol.

  15. Scalable air cathode microbial fuel cells using glass fiber separators, plastic mesh supporters, and graphite fiber brush anodes

    KAUST Repository

    Zhang, Xiaoyuan; Cheng, Shaoan; Liang, Peng; Huang, Xia; Logan, Bruce E.

    2011-01-01

    The combined use of brush anodes and glass fiber (GF1) separators, and plastic mesh supporters were used here for the first time to create a scalable microbial fuel cell architecture. Separators prevented short circuiting of closely

  16. Fully Aromatic Block Copolymers for Fuel Cell Membranes with Densely Sulfonated Nanophase Domains

    DEFF Research Database (Denmark)

    Takamuku, Shogo; Jannasch, Patrick; Lund, Peter Brilner

    Two multiblock copoly(arylene ether sulfone)s with similar block lengths and ion exchange capacities (IECs) were prepared by a coupling reaction between a non-sulfonated precursor block and a highly sulfonated precursor block containing either fully disulfonated diarylsulfone or fully...... tetrasulfonated tetraaryldisulfone segments. The latter two precursor blocks were sulfonated via lithiation-sulfination reactions whereby the sulfonic acid groups were exclu- sively placed in ortho positions to the many sulfone bridges, giving these locks IECs of 4.1 and 4.6 meqg1, respectively. Copolymer...

  17. Fuel behavior and fission product release under HTGR accident conditions

    International Nuclear Information System (INIS)

    Fukuda, K.; Hayashi, K.; Shiba, K.

    1990-01-01

    In early 1989 a final decision was made over construction of a 30 MWth HTGR called the High Temperature Engineering Test Reactor, HTTR, in Japan in order to utilize it for high temperature gas engineering tests and various nuclear material tests. The HTTR fuel is a pin-in-block type fuel element which is composed of a hexagonal graphite block with dimension of 580 mm in length and 360 mm in face-to-face distance and about 30 of the fuel rods inserted into the coolant channels drilled in the block. The TRISO coated fuel particles for HTTR are incorporated with graphite powder and phenol resin into the fuel compacts, 19 of which are encased into a graphite sleeve as a fuel rod. It is necessary for the HTTR licensing to prove the fuel stability under predicted accidents related to the high temperature events. Therefore, the release of the fission products and the fuel failure have been investigated in the irradiation---and the heating experiments simulating these conditions at JAERI. This report describes the HTTR fuel behavior at extreme temperature, made clear in these experiments

  18. 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

  19. Block copolymer based composition and morphology control in nanostructured hybrid materials for energy conversion and storage: solar cells, batteries, and fuel cells

    KAUST Repository

    Orilall, M. Christopher; Wiesner, Ulrich

    2011-01-01

    to materials design and current limitations, the review will highlight some of the most recent examples of block copolymer structure-directed nanomaterials for photovoltaics, batteries and fuel cells. In each case insights are provided into the various

  20. In depth fusion flame spreading with a deuterium—tritium plane fuel density profile for plasma block ignition

    International Nuclear Information System (INIS)

    Malekynia, B.; Razavipour, S. S.

    2012-01-01

    Solid-state fuel ignition was given by Chu and Bobin according to the hydrodynamic theory at x = 0 qualitatively. A high threshold energy flux density, i.e., E* = 4.3 × 10 12 J/m 2 , has been reached. Recently, fast ignition by employing clean petawatt—picosecond laser pulses was performed. The anomalous phenomena were observed to be based on suppression of prepulses. The accelerated plasma block was used to ignite deuterium—tritium fuel at solid-state density. The detailed analysis of the thermonuclear wave propagation was investigated. Also the fusion conditions at x ≠ 0 layers were clarified by exactly solving hydrodynamic equations for plasma block ignition. In this paper, the applied physical mechanisms are determined for nonlinear force laser driven plasma blocks, thermonuclear reaction, heat transfer, electron—ion equilibration, stopping power of alpha particles, bremsstrahlung, expansion, density dependence, and fluid dynamics. New ignition conditions may be obtained by using temperature equations, including the density profile that is obtained by the continuity equation and expansion velocity. The density is only a function of x and independent of time. The ignition energy flux density, E* t , for the x ≠ 0 layers is 1.95 × 10 12 J/m 2 . Thus threshold ignition energy in comparison with that at x = 0 layers would be reduced to less than 50 percent. (physics of gases, plasmas, and electric discharges)

  1. 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

  2. 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.

  3. A Graphite Isotope Ratio Method: A Primer on Estimating Plutonium Production in Graphite Moderated Reactors

    International Nuclear Information System (INIS)

    Gesh, Christopher J.

    2004-01-01

    The Graphite Isotope Ratio Method (GIRM) is a technique used to estimate the total plutonium production in a graphite-moderated reactor. The cumulative plutonium production in that reactor can be accurately determined by measuring neutron irradiation induced isotopic ratio changes in certain impurity elements within the graphite moderator. The method does not require detailed knowledge of a reactor's operating history, although that knowledge can decrease the uncertainty of the production estimate. The basic premise of the Graphite Isotope Ratio Method is that the fluence in non-fuel core components is directly related to the cumulative plutonium production in the nuclear fuel

  4. Evaluation of the oxidation behavior and strength of the graphite components in the VHTR, (1)

    International Nuclear Information System (INIS)

    Eto, Motokuni; Kurosawa, Takeshi; Nomura, Shinzo; Imai, Hisashi

    1987-04-01

    Oxidation experiments have been carried out mainly on a fine-grained isotropic graphite, IG-110, at temperatures between 1173 and 1473 K in a water vapor/helium mixture. In most cases water vapor concentration was 0.65 vol% and helium pressure, 1 atm. Reaction rate and burn-off profile were measured using cylindrical specimens. On the basis of the experimental data the oxidation behavior of fuel block and core support post under the condition of the VHTR operation was estimated using the first-order or Langmuir-Hinshelwood equation with regard to water vapor concentration. Strength and stress-strain relationship of the graphite components with burn-off profiles estimated above were analyzed on the basis of the model for stress-strain relationship and strength of graphite specimens with density gradients. The estimation indicated that the integrity of the components would be maintained during normal reactor operation. (author)

  5. Investigation on structural integrity of graphite component during high temperature 950degC continuous operation of HTTR

    International Nuclear Information System (INIS)

    Sumita, Junya; Shimazaki, Yosuke; Shibata, Taiju

    2014-01-01

    Graphite material is used for internal structures in high temperature gas-cooled reactor. The core components and graphite core support structures are so designed as to maintain the structural integrity to keep core cooling capability. To confirm that the core components and graphite core support structures satisfy the design requirements, the temperatures of the reactor internals are measured during the reactor operation. Surveillance test of graphite specimens and in-service inspection using TV camera are planned in conjunction with the refueling. This paper describes the evaluation results of the integrity of the core components and graphite core support structures during the high temperature 950degC continuous operation, a high temperature continuous operation with reactor outlet temperature of 950degC for 50 days, in high temperature engineering test reactor. The design requirements of the core components and graphite core support structures were satisfied during the high temperature 950degC continuous operation. The dimensional change of graphite which directly influences the temperature of coolant was estimated considering the temperature profiles of fuel block. The magnitude of irradiation-induced dimensional change considering temperature profiles was about 1.2 times larger than that under constant irradiation temperature of 1000degC. In addition, the programs of surveillance test and ISI using TV camera were introduced. (author)

  6. Deoxyribonucleic acid directed metallization of platinum nanoparticles on graphite nanofibers as a durable oxygen reduction catalyst for polymer electrolyte fuel cells

    Science.gov (United States)

    Peera, S. Gouse; Sahu, A. K.; Arunchander, A.; Nath, Krishna; Bhat, S. D.

    2015-11-01

    Effective surface functionalization to the hydrophobic graphite nanofibers (GNF) is performed with the biomolecule, namely deoxy-ribo-nucleic-acid (DNA) via π-π interactions. Pt nanoparticles are impregnated on GNF-DNA composite by ethylene glycol reduction method (Pt/GNF-DNA) and its effect on electro catalytic activity for oxygen reduction reaction (ORR) is systemically studied. Excellent dispersion of Pt nanoparticles over GNF-DNA surfaces with no evidence on particle aggregation is a remarkable achievement in this study. This result in higher electro chemical surface area of the catalyst, enhanced ORR behavior with significant enhancement in mass activity. The catalyst is validated in H2-O2 polymer electrolyte fuel cell (PEFC) and a peak power density of 675 mW cm-2 is achieved at a load current density of 1320 mA cm-2 with a minimal catalyst loading of 0.1 mg cm-2 at a cell temperature of 70 °C and 2 bar absolute pressure. Repeated potential cycling up to 10000 cycles in acidic media is also performed for this catalyst and found excellent stability with only 60 mV drop in the ORR half wave potential. The superior behavior of Pt/GNF-DNA catalyst is credited to the robust fibrous structure of GNF and its effective surface functionalization process via π-π interaction.

  7. 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.

  8. Aseismic study of high temperature gas-cooled reactor core with block-type fuel, 3

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1985-01-01

    A two-dimensional horizontal seismic experiment with single axis and simultaneous two-axes excitations was performed to obtain the core seismic design data on the block-type high temperature gas-cooled reactor. Effects of excitation directions and core side support stiffness on characteristics of core displacements and reaction forces of support were revealed. The values of the side reaction forces are the largest in the excitation of flat-to-flat of hexagonal block. Preload from the core periphery to the core center are effective to decrease core displacements and side reaction forces. (author)

  9. Fluorinated building blocks for next-generation polymer electrolyte membrane fuel cells

    NARCIS (Netherlands)

    Wadekar, M.N.

    2012-01-01

    The purpose of this thesis is to design, create and study basic building blocks for the construction of self-assembled nanostructured electrodes and membranes for PEMFC. The research described deals with the synthesis of polymerizable fluorosurfactant (1) and its non-polymerizable analogue (2) and

  10. Polybenzimidazole block copolymers for fuel cell: synthesis and studies of block length effects on nanophase separation, mechanical properties, and proton conductivity of PEM.

    Science.gov (United States)

    Maity, Sudhangshu; Jana, Tushar

    2014-05-14

    A series of meta-polybenzimidazole-block-para-polybenzimidazole (m-PBI-b-p-PBI), segmented block copolymers of PBI, were synthesized with various structural motifs and block lengths by condensing the diamine terminated meta-PBI (m-PBI-Am) and acid terminated para-PBI (p-PBI-Ac) oligomers. NMR studies and existence of two distinct glass transition temperatures (Tg), obtained from dynamical mechanical analysis (DMA) results, unequivocally confirmed the formation of block copolymer structure through the current polymerization methodology. Appropriate and careful selection of oligomers chain length enabled us to tailor the block length of block copolymers and also to make varieties of structural motifs. Increasingly distinct Tg peaks with higher block length of segmented block structure attributed the decrease in phase mixing between the meta-PBI and para-PBI blocks, which in turn resulted into nanophase segregated domains. The proton conductivities of proton exchange membrane (PEM) developed from phosphoric acid (PA) doped block copolymer membranes were found to be increasing substantially with increasing block length of copolymers even though PA loading of these membranes did not alter appreciably with varying block length. For example when molecular weight (Mn) of blocks were increased from 1000 to 5500 then the proton conductivities at 160 °C of resulting copolymers increased from 0.05 to 0.11 S/cm. Higher block length induced nanophase separation between the blocks by creating less morphological barrier within the block which facilitated the movement of the proton in the block and hence resulting higher proton conductivity of the PEM. The structural varieties also influenced the phase separation and proton conductivity. In comparison to meta-para random copolymers reported earlier, the current meta-para segmented block copolymers were found to be more suitable for PBI-based PEM.

  11. Scalable air cathode microbial fuel cells using glass fiber separators, plastic mesh supporters, and graphite fiber brush anodes

    KAUST Repository

    Zhang, Xiaoyuan

    2011-01-01

    The combined use of brush anodes and glass fiber (GF1) separators, and plastic mesh supporters were used here for the first time to create a scalable microbial fuel cell architecture. Separators prevented short circuiting of closely-spaced electrodes, and cathode supporters were used to avoid water gaps between the separator and cathode that can reduce power production. The maximum power density with a separator and supporter and a single cathode was 75±1W/m3. Removing the separator decreased power by 8%. Adding a second cathode increased power to 154±1W/m3. Current was increased by connecting two MFCs connected in parallel. These results show that brush anodes, combined with a glass fiber separator and a plastic mesh supporter, produce a useful MFC architecture that is inherently scalable due to good insulation between the electrodes and a compact architecture. © 2010 Elsevier Ltd.

  12. 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.

  13. 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

  14. Treat upgrade fuel fabrication

    International Nuclear Information System (INIS)

    Davidson, K.V.; Schell, D.H.

    1979-01-01

    An extrusion and thermal treatment process was developed to produce graphite fuel rods containing a dispersion of enriched UO 2 . These rods will be used in an upgraded version of the Transient Reactor Test Facility (TREAT). The improved fuel provides a higher graphite matrix density, better fuel dispersion and higher thermal capabilities than the existing fuel

  15. 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)

  16. 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.

  17. Thermal-hydraulic of partially blocked fuel subassembly with porous media

    International Nuclear Information System (INIS)

    Nagata, Takemitsu; Ohshima, Hiroyuki

    2000-10-01

    The analysis code for investigations of local subassembly phenomena, which has been recognized as an issue of local subassembly accidents, has been required and developed at JNC. It is desirable for the analysis code to be applicable to various blockage conditions and random position of the blockage formation and to evaluate conservatively on the safety assessment with high accuracy. In this study, for the purpose of verifying the application and issues of the subchannel analysis code ASFRE-IV which evaluates thermal hydraulic phenomena in the porous blockage regions, the ASFRE-IV validation analysis was carried out on the basis of the data of an experiment investigation on a local porous blockage in a fuel subassembly performed by Reactor Engineering Groop, O-arai Engineering Center, JNC. Calculational results indicated that ASFRE-IV could reproduce the coolant temperature profile in a fuel subassembly and the peak temperature in the local subchannel conservatively. (author)

  18. 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

  19. 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)

  20. Imidazolium-based Block Copolymers as Solid-State Separators for Alkaline Fuel Cells and Lithium Ion Batteries

    Science.gov (United States)

    Nykaza, Jacob Richard

    In this study, polymerized ionic liquid (PIL) diblock copolymers were explored as solid-state polymer separators as an anion exchange membrane (AEM) for alkaline fuel cells AFCs and as a solid polymer electrolyte (SPE) for lithium-ion batteries. Polymerized ionic liquid (PIL) block copolymers are a distinct set of block copolymers that combine the properties of both ionic liquids (e.g., high conductivity, high electrochemical stability) and block copolymers (e.g., self-assembly into various nanostructures), which provides the opportunity to design highly conductive robust solid-state electrolytes that can be tuned for various applications including AFCs and lithium-ion batteries via simple anion exchange. A series of bromide conducting PIL diblock copolymers with an undecyl alkyl side chain between the polymer backbone and the imidazolium moiety were first synthesized at various compositions comprising of a PIL component and a non-ionic component. Synthesis was achieved by post-functionalization from its non-ionic precursor PIL diblock copolymer, which was synthesized via the reverse addition fragmentation chain transfer (RAFT) technique. This PIL diblock copolymer with long alkyl side chains resulted in flexible, transparent films with high mechanical strength and high bromide ion conductivity. The conductivity of the PIL diblock copolymer was three times higher than its analogous PIL homopolymer and an order of magnitude higher than a similar PIL diblock copolymer with shorter alkyl side chain length, which was due to the microphase separated morphology, more specifically, water/ion clusters within the PIL microdomains in the hydrated state. Due to the high conductivity and mechanical robustness of this novel PIL block copolymer, its application as both the ionomer and AEM in an AFC was investigated via anion exchange to hydroxide (OH-), where a maximum power density of 29.3 mW cm-1 (60 °C with H2/O2 at 25 psig (172 kPa) backpressure) was achieved. Rotating disk

  1. 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).

  2. Polyarylenethioethersulfone Membranes for Fuel Cells (Postprint)

    Science.gov (United States)

    2010-01-01

    The Electrochemical SocietyProton exchange membrane fuel cells PEMFCs are an attrac- tive power source due to their energy efficiency and...standard in PEMFC technology.3,4 Nafion membranes have a polytetrafluoro- ethylene PTFE backbone, which provides thermal and chemical stability, and...diffusion layers to fabricate MEAs. Single-cell test (H- PEMFC ).— MEAs were positioned in a single-cell fixture with graphite blocks as current

  3. Vapour pressure of caesium over nuclear graphite

    International Nuclear Information System (INIS)

    Faircloth, R.L.; Pummery, F.C.W.

    1976-01-01

    The vapour pressure of caesium over a fine-grained isotropic moulded gilsocarbon nuclear graphite intended for use in the manufacture of fuel tubes for the high temperature reactor has been determined as a function of temperature and concentration by means of the Knudsen effusion technique. The concentration range 0 to 10 μg caesium/g graphite was investigated and it was concluded that a Langmuir adsorption situation exists under these conditions. (author)

  4. Determination of a geometry-dependent parameter and development of a calculation model for describing the fission products transport from spherical fuel elements of graphite moderated gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Weissfloch, R

    1973-07-15

    The fuel elements of high-temperature reactors, coated with pyrolitic carbon and covered with graphite, release fission products like all other fuel elements. Because of safety reasons, the rate of this release has to be kept low and has also to be predictable. Measured values from irradiation tests and from post-irradiation tests about the actual release of different fission products are presented. The physical and chemical mechanism, which determines the release, is extraordinarily complex and in particular not clearly defined. Because of the mentioned reasons, a simplified calculation model was developed, which only considers the release-mechanisms phenomenologically. This calculation model coincides very well in its results with values received in experiments until now. It can be held as an interim state on the way to a complete theory.

  5. Determination of a geometry-dependent parameter and development of a calculation model for describing the fission products transport from spherical fuel elements of graphite moderated gas-cooled reactors

    International Nuclear Information System (INIS)

    Weissfloch, R.

    The fuel elements of High-Temperature Reactors, coated with pyrolitic carbon and covered with graphite, release fission products like all other fuel elements. Because of safety reasons the rate of this release has to be kept low and has also to be predictable. Measured values from irradiation tests and from post-irradiation tests about the actual release of different fission products are present. The physical and chemical mechanism, which determines the release, is extraordinarily complex and in particular not clearly defined. Because of the mentioned reasons a simplified calculation model was developed, which only considers the release-mechanisms phenomenologically. This calculation model coincides very well in its results with values received in experiments until now. It can serve as an interim state on the way to a complete theory. (U.S.)

  6. Progress on the development of a new fuel management code to simulate the movement of pebble and block type fuel elements in a very high temperature reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Xhonneux, Andre, E-mail: a.xhonneux@fz-juelich.de [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology, RWTH-Aachen, 52064 Aachen (Germany); Kasselmann, Stefan; Rütten, Hans-Jochem [Forschungszentrum Jülich, 52425 Jülich (Germany); Becker, Kai [Institute for Reactor Safety and Reactor Technology, RWTH-Aachen, 52064 Aachen (Germany); Allelein, Hans-Josef [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology, RWTH-Aachen, 52064 Aachen (Germany)

    2014-05-01

    The history of gas-cooled high-temperature reactor prototypes in Germany is closely related to Forschungszentrum Jülich and its “Institute of Nuclear Waste Disposal and Reactor Safety (IEK-6)”. A variety of computer codes have been developed, validated and optimized to simulate the different safety and operational aspects of V/HTR. In order to overcome the present limitations of these codes and to exploit the advantages of modern computer clusters, a project has been initiated to integrate these individual programs into a consistent V/HTR code package (VHCP) applying state-of-the-art programming techniques and standards. One important aspect in the simulation of a V/HTR is the modeling of a continuous moving pebble bed or the periodic rearrangement of prismatic block type fuel. Present models are either too coarse to take special issues (e.g. pebble piles) into account or are too detailed and therefore too time consuming to be applicable in the HCP. The new Software for Handling Universal Fuel Elements (SHUFLE) recently being developed is well suited to close this gap. Although at first the code has been designed for pebble bed reactors, it can in principal be applied to all other types of nuclear fuel. The granularity of the mesh grid meets the requirements to consider these special issues while keeping the used computing power within reasonable limits. New features are for example the possibility to consider azimuthally differing flow velocities in the case of a pebble bed reactor or individual void factors to simulate effects to seismic events. The general idea behind this new approach to the simulation of pebble bed reactors is the following: In the preprocessing step, experimental flow lines or flow lines simulated by more detailed codes serve as an input. For each radial mesh column a representative flow line is then determined by interpolation. These representative flow lines are finally mapped to a user defined rectangular grid forming chains of meshes

  7. Progress on the development of a new fuel management code to simulate the movement of pebble and block type fuel elements in a very high temperature reactor core

    International Nuclear Information System (INIS)

    Xhonneux, Andre; Kasselmann, Stefan; Rütten, Hans-Jochem; Becker, Kai; Allelein, Hans-Josef

    2014-01-01

    The history of gas-cooled high-temperature reactor prototypes in Germany is closely related to Forschungszentrum Jülich and its “Institute of Nuclear Waste Disposal and Reactor Safety (IEK-6)”. A variety of computer codes have been developed, validated and optimized to simulate the different safety and operational aspects of V/HTR. In order to overcome the present limitations of these codes and to exploit the advantages of modern computer clusters, a project has been initiated to integrate these individual programs into a consistent V/HTR code package (VHCP) applying state-of-the-art programming techniques and standards. One important aspect in the simulation of a V/HTR is the modeling of a continuous moving pebble bed or the periodic rearrangement of prismatic block type fuel. Present models are either too coarse to take special issues (e.g. pebble piles) into account or are too detailed and therefore too time consuming to be applicable in the HCP. The new Software for Handling Universal Fuel Elements (SHUFLE) recently being developed is well suited to close this gap. Although at first the code has been designed for pebble bed reactors, it can in principal be applied to all other types of nuclear fuel. The granularity of the mesh grid meets the requirements to consider these special issues while keeping the used computing power within reasonable limits. New features are for example the possibility to consider azimuthally differing flow velocities in the case of a pebble bed reactor or individual void factors to simulate effects to seismic events. The general idea behind this new approach to the simulation of pebble bed reactors is the following: In the preprocessing step, experimental flow lines or flow lines simulated by more detailed codes serve as an input. For each radial mesh column a representative flow line is then determined by interpolation. These representative flow lines are finally mapped to a user defined rectangular grid forming chains of meshes

  8. Numerical simulations of helium flow through prismatic fuel elements of very high temperature reactors

    International Nuclear Information System (INIS)

    Ribeiro, Felipe Lopes; Pinto, Joao Pedro C.T.A.

    2013-01-01

    The 4 th generation Very High Temperature Reactor (VHTR) most popular concept uses a graphite-moderated and helium cooled core with an outlet gas temperature of approximately 1000 deg C. The high output temperature allows the use of the process heat and the production of hydrogen through the thermochemical iodine-sulfur process as well as highly efficient electricity generation. There are two concepts of VHTR core: the prismatic block and the pebble bed core. The prismatic block core has two popular concepts for the fuel element: multihole and annular. In the multi-hole fuel element, prismatic graphite blocks contain cylindrical flow channels where the helium coolant flows removing heat from cylindrical fuel rods positioned in the graphite. In the other hand, the annular type fuel element has annular channels around the fuel. This paper shows the numerical evaluations of prismatic multi-hole and annular VHTR fuel elements and does a comparison between the results of these assembly reactors. In this study the analysis were performed using the CFD code ANSYS CFX 14.0. The simulations were made in 1/12 fuel element models. A numerical validation was performed through the energy balance, where the theoretical and the numerical generated heat were compared for each model. (author)

  9. Graphite Oxidation Thermodynamics/Reactions

    International Nuclear Information System (INIS)

    Propp, W.A.

    1998-01-01

    The vulnerability of graphite-matrix spent nuclear fuel to oxidation by the ambient atmosphere if the fuel canister is breached was evaluated. Thermochemical and kinetic data over the anticipated range of storage temperatures (200 to 400 C) were used to calculate the times required for a total carbon mass loss of 1 mgcm-2 from a fuel specimen. At 200 C, the time required to produce even this small loss is large, 900,000 yr. However, at 400 C the time required is only 1.9 yr. The rate of oxidation at 200 C is negligible, and the rate even at 400 C is so small as to be of no practical consequence. Therefore, oxidation of the spent nuclear fuel upon a loss of canister integrity is not anticipated to be a concern based upon the results of this study

  10. Observation of Compressive Deformation Behavior of Nuclear Graphite by Digital Image Correlation

    International Nuclear Information System (INIS)

    Kim, Hyunju; Kim, Eungseon; Kim, Minhwan; Kim, Yongwan

    2014-01-01

    Polycrystalline nuclear graphite has been proposed as a fuel element, moderator and reflector blocks, and core support structures in a very high temperature gas-cooled reactor. During reactor operation, graphite core components and core support structures are subjected to various stresses. It is therefore important to understand the mechanism of deformation and fracture of nuclear graphites, and their significance to structural integrity assessment methods. Digital image correlation (DIC) is a powerful tool to measure the full field displacement distribution on the surface of the specimens. In this study, to gain an understanding of compressive deformation characteristic, the formation of strain field during a compression test was examined using a commercial DIC system. An examination was made to characterize the compressive deformation behavior of nuclear graphite by a digital image correlation. The non-linear load-displacement characteristic prior to the peak load was shown to be mainly dominated by the presence of localized strains, which resulted in a permanent displacement. Young's modulus was properly calculated from the measured strain

  11. 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

  12. Study on "1"4C content in post-irradiation graphite spheres of HTR-10

    International Nuclear Information System (INIS)

    Wang Shouang; Pi Yue; Xie Feng; Li Hong; Cao Jianzhu

    2014-01-01

    Since the production mechanism of the "1"4C in spherical fuel elements was similar to that of fuel-free graphite spheres, in order to obtain the amount of "1"4C in fuel elements and graphite spheres of HTR-10, the production mechanism of the "1"4C in graphite spheres was studied. The production sources of the "1"4C in graphite spheres and fuel elements were summarized, the amount of "1"4C in the post-irradiation graphite spheres was calculated, the decomposition techniques of graphite spheres were compared, and experimental methods for decomposing the graphite spheres and preparing the "1"4C sample were proposed. The results can lay the foundation for further experimental research and provide theoretical calculations for comparison. (authors)

  13. Graphite moderated 252Cf source

    International Nuclear Information System (INIS)

    Sajo B, L.; Barros, H.; Greaves, E. D.; Vega C, H. R.

    2014-08-01

    The thorium molten salt reactor is an attractive and affordable nuclear power option for developing countries with insufficient infrastructure and limited technological capability. In the aim of personnel training and experience gathering at the Universidad Simon Bolivar there is in progress a project of developing a subcritical thorium liquid fuel reactor. The neutron source to run this subcritical reactor is a 252 Cf source and the reactor will use high-purity graphite as moderator. Using the MCNP5 code the neutron spectra of the 252 Cf in the center of the graphite moderator has been estimated along the channel where the liquid thorium salt will be inserted; also the ambient dose equivalent due to the source has been determined around the moderator. (Author)

  14. 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

  15. 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.

  16. The internal propagation of fusion flame with the strong shock of a laser driven plasma block for advanced nuclear fuel ignition

    International Nuclear Information System (INIS)

    Malekynia, B.; Razavipour, S. S.

    2013-01-01

    An accelerated skin layer may be used to ignite solid state fuels. Detailed analyses were clarified by solving the hydrodynamic equations for nonlinear force driven plasma block ignition. In this paper, the complementary mechanisms are included for the advanced fuel ignition: external factors such as lasers, compression, shock waves, and sparks. The other category is created within the plasma fusion as reheating of an alpha particle, the Bremsstrahlung absorption, expansion, conduction, and shock waves generated by explosions. With the new condition for the control of shock waves, the spherical deuterium-tritium fuel density should be increased to 75 times that of the solid state. The threshold ignition energy flux density for advanced fuel ignition may be obtained using temperature equations, including the ones for the density profile obtained through the continuity equation and the expansion velocity for the r ≠ 0 layers. These thresholds are significantly reduced in comparison with the ignition thresholds at x = 0 for solid advanced fuels. The quantum correction for the collision frequency is applied in the case of the delay in ion heating. Under the shock wave condition, the spherical proton-boron and proton-lithium fuel densities should be increased to densities 120 and 180 times that of the solid state. These plasma compressions are achieved through a longer duration laser pulse or X-ray. (physics of gases, plasmas, and electric discharges)

  17. The investigation of HTGR fuel regeneration process

    Energy Technology Data Exchange (ETDEWEB)

    Lazarev, L N; Bertina, L E; Popik, V P; Isakov, V P; Alkhimov, N B; Pokhitonov, Yu A

    1985-07-01

    The aim of this report is the investigation of HTGR fuel regeneration. The operation in the technologic scheme of uranium extraction from fuel depleted elements is separation of fuel from graphite. Available methods of graphite matrix destruction are: mechanical destruction, chemical destruction, and burning. Mechanical destruction is done in combination with leaching or chlorination. Methods of chemical destruction of graphite matrix are not sufficiently studied. Most of the investigations nowadays sre devoted to removal of graphite by burning.

  18. The investigation of HTGR fuel regeneration process

    International Nuclear Information System (INIS)

    Lazarev, L.N.; Bertina, L.E.; Popik, V.P.; Isakov, V.P.; Alkhimov, N.B.; Pokhitonov, Yu.A.

    1985-01-01

    The aim of this report is the investigation of HTGR fuel regeneration. The operation in the technologic scheme of uranium extraction from fuel depleted elements is separation of fuel from graphite. Available methods of graphite matrix destruction are: mechanical destruction, chemical destruction, and burning. Mechanical destruction is done in combination with leaching or chlorination. Methods of chemical destruction of graphite matrix are not sufficiently studied. Most of the investigations nowadays sre devoted to removal of graphite by burning

  19. The roles of geometry and topology structures of graphite fillers on thermal conductivity of the graphite/aluminum composites

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, C.; Chen, D.; Zhang, X.B. [State Key Laboratory of Metal Matrix Composites, Shanghai Jiao Tong University, Shanghai 200240 (China); Chen, Z., E-mail: zhe.chen@sjtu.edu.cn [State Key Laboratory of Metal Matrix Composites, Shanghai Jiao Tong University, Shanghai 200240 (China); Zhong, S.Y.; Wu, Y. [State Key Laboratory of Metal Matrix Composites, Shanghai Jiao Tong University, Shanghai 200240 (China); Ji, G. [Unité Matériaux et Transformations, CNRS UMR 8207, Université Lille 1, Villeneuve d' Ascq 59655 (France); Wang, H.W. [State Key Laboratory of Metal Matrix Composites, Shanghai Jiao Tong University, Shanghai 200240 (China)

    2015-02-20

    Various graphite fillers, such as graphite particles, graphite fibers, graphite flakes and porous graphite blocks, have been successfully incorporated into an Al alloy by squeeze casting in order to fabricate graphite/Al composites with enhanced thermal conductivity (TC). Microstructural characterization by X-ray diffraction and scanning electron microscopy has revealed a tightly-adhered, clean and Al{sub 4}C{sub 3}-free interface between the graphite fillers and the Al matrix in all the as-fabricated composites. Taking the microstructural features into account, we generalized the corresponding predictive models for the TCs of these composites with the effective medium approximation and the Maxwell mean-field scheme, which both show good agreement with the experimental data. The roles of geometry and topology structures of graphite fillers on the TCs of the composites were further discussed. - Highlights: • The thermal enhancement of various graphite fillers with different topology structures. • Predictive models for the thermal conductivity of different topology structures. • Oriented flakes alignment has the high potentials for thermal enhancement.

  20. Hydrogen storage in graphitic nanofibres

    OpenAIRE

    McCaldin, Simon Roger

    2007-01-01

    There is huge need to develop an alternative to hydrocarbons fuel, which does not produce CO2 or contribute to global warming - 'the hydrogen economy' is such an alternative, however the storage of hydrogen is the key technical barrier that must be overcome. The potential of graphitic nanofibres (GNFs) to be used as materials to allow the solid-state storage of hydrogen has thus been investigated. This has been conducted with a view to further developing the understanding of the mechanism(s) ...

  1. 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

  2. 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....

  3. 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.

  4. 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.

  5. Misorientations in spheroidal graphite: some new insights about spheroidal graphite growth in cast irons

    International Nuclear Information System (INIS)

    Lacaze, J; Theuwissen, K; Laffont, L; Véron, M

    2016-01-01

    Local diffraction patterning, orientation mapping and high resolution transmission electron microscopy imaging have been used to characterize misorientations in graphite spheroids of cast irons. Emphasis is put here on bulk graphite, away from the nucleus as well as from the outer surface of the spheroids in order to get information on their growth during solidification. The results show that spheroidal graphite consists in conical sectors made of elementary blocks piled up on each other. These blocks are elongated along the prismatic a direction of graphite with the c axes roughly parallel to the radius of the spheroids. This implies that the orientation of the blocks rotates around the spheroid centre giving low angle tilting misorientations along tangential direction within each sector. Misorientations between neighbouring sectors are of higher values and their interfaces show rippled layers which are characteristic of defects in graphene. Along a radius of the spheroid, clockwise and anticlockwise twisting between blocks is observed. These observations help challenging some of the models proposed to explain spheroidal growth in cast ions. (paper)

  6. 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

  7. 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.

  8. 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

  9. Internal reforming of methane in a mono-block-layer build solid oxide fuel cell with an embedding porous pipe: Numerical analysis

    International Nuclear Information System (INIS)

    Ramírez-Minguela, J.J.; Mendoza-Miranda, J.M.; Muñoz-Carpio, V.D.; Rangel-Hernández, V.H.; Pérez-García, V.; Rodríguez-Muñoz, J.L.

    2014-01-01

    Highlights: • A MOLB-type SOFC with different biogas compositions is modeled. • Internal reforming of methane and water gas shift in the anode side are considered. • Changes in the SOFC according to the phenomena inside of them are assessed. • A new configuration for the MOLB-type geometry is proposed. • Temperature gradients effects in the new configuration are significant. - Abstract: A mono-block-layer build type (MOLB-type) geometry, composed with trapezoidal cross-section channels for fuel and air, of solid oxide fuel cell (SOFC), is modeled in this study using a three dimensional computational fluid dynamics (CFD). The model is used to obtain the performance of the SOFC considering internal reforming of methane in the anode side with different biogas compositions. From the results obtained, a new configuration for the MOLB-type geometry is proposed. The model takes into account the mass transfer, heat transfer, species transport, chemical and electrochemical reactions, detailed comparisons of species concentration, temperature and electric fields inside the cell analyzed in the present study. Results show that the biogas from local sludge has the lower temperature gradient and more homogeneous current density distributions along of the fuel cell, on the other hand, with the new configuration proposed for the MOLB-type geometry the temperature distribution inside of the fuel cell has lower temperature gradients

  10. Multiblock copolymers with highly sulfonated blocks containing di- and tetrasulfonated arylene sulfone segments for proton exchange membrane fuel cell applications

    Energy Technology Data Exchange (ETDEWEB)

    Takamuku, Shogo; Jannasch, Patric [Polymer and Materials Chemistry, Department of Chemistry, Lund University (Sweden)

    2012-01-15

    Multiblock copoly(arylene ether sulfone)s with different block lengths and ionic contents are tailored for durable and proton-conducting electrolyte membranes. Two series of fully aromatic copolymers are prepared by coupling reactions between non-sulfonated hydrophobic precursor blocks and highly sulfonated hydrophilic precursor blocks containing either fully disulfonated diarylsulfone or fully tetrasulfonated tetraaryldisulfone segments. The sulfonic acid groups are exclusively introduced in ortho positions to the sulfone bridges to impede desulfonation reactions and give the blocks ion exchange capacities (IECs) of 4.1 and 4.6 meq. g{sup -1}, respectively. Solvent cast block copolymer membranes show well-connected hydrophilic nanophase domains for proton transport and high decomposition temperatures above 310 C under air. Despite higher IEC values, membranes containing tetrasulfonated tetraaryldisulfone segments display a markedly lower water uptake than the corresponding ones with disulfonated diarylsulfone segments when immersed in water at 100 C, presumably because of the much higher chain stiffness and glass transition temperature of the former segments. The former membranes have proton conductivities in level of a perfluorosulfonic acid membrane (NRE212) under fully humidified conditions. A membrane with an IEC of 1.83 meq. g{sup -1} reaches above 6 mS cm{sup -1} under 30% relative humidity at 80 C, to be compared with 10 mS cm{sup -1} for NRE212 under the same conditions. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  11. 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.

  12. 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

  13. CFD Analysis of the Fuel Temperature in High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    In, W. K.; Chun, T. H.; Lee, W. J.; Chang, J. H.

    2005-01-01

    High temperature gas-cooled reactors (HTGR) have received a renewed interest as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor (PBR) and a prismatic modular reactor (PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both PBR and PMR. The objective of this study is to predict the fuel temperature distributions in PBR and PMR using a computational fluid dynamics(CFD) code, CFX-5. The reference reactor designs used in this analysis are PBMR400 and GT-MHR600

  14. Development of integrated waste management options for irradiated graphite

    Directory of Open Access Journals (Sweden)

    Alan Wareing

    2017-08-01

    Full Text Available The European Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project sought to develop best practices in the retrieval, treatment, and disposal of irradiated graphite including other irradiated carbonaceous waste such as structural material made of graphite, nongraphitized carbon bricks, and fuel coatings. Emphasis was given on legacy irradiated graphite, as this represents a significant inventory in respective national waste management programs. This paper provides an overview of the characteristics of graphite irradiated during its use, primarily as a moderator material, within nuclear reactors. It describes the potential techniques applicable to the retrieval, treatment, recycling/reuse, and disposal of these graphite wastes. Considering the lifecycle of nuclear graphite, from manufacture to final disposal, a number of waste management options have been developed. These options consider the techniques and technologies required to address each stage of the lifecycle, such as segregation, treatment, recycle, and ultimate disposal in a radioactive waste repository, providing a toolbox to aid operators and regulators to determine the most appropriate management strategy. It is noted that national waste management programs currently have, or are in the process of developing, respective approaches to irradiated graphite management. The output of the Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project is intended to aid these considerations, rather than dictate them.

  15. Development of integrated waste management options for irradiated graphite

    Energy Technology Data Exchange (ETDEWEB)

    Wareing, Alan; Abrahamsen-Mills, Liam; Fowler, Linda; Jarvis, Richard; Banford, Anthony William [National Nuclear Laboratory, Warrington (United Kingdom); Grave, Michael [Doosan Babcock, Gateshead (United Kingdom); Metcalfe, Martin [National Nuclear Laboratory, Gloucestershire (United Kingdom); Norris, Simon [Radioactive Waste Management Limited, Oxon (United Kingdom)

    2017-08-15

    The European Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project sought to develop best practices in the retrieval, treatment, and disposal of irradiated graphite including other irradiated carbonaceous waste such as structural material made of graphite, nongraphitized carbon bricks, and fuel coatings. Emphasis was given on legacy irradiated graphite, as this represents a significant inventory in respective national waste management programs. This paper provides an overview of the characteristics of graphite irradiated during its use, primarily as a moderator material, within nuclear reactors. It describes the potential techniques applicable to the retrieval, treatment, recycling/reuse, and disposal of these graphite wastes. Considering the lifecycle of nuclear graphite, from manufacture to final disposal, a number of waste management options have been developed. These options consider the techniques and technologies required to address each stage of the lifecycle, such as segregation, treatment, recycle, and ultimate disposal in a radioactive waste repository, providing a toolbox to aid operators and regulators to determine the most appropriate management strategy. It is noted that national waste management programs currently have, or are in the process of developing, respective approaches to irradiated graphite management. The output of the Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project is intended to aid these considerations, rather than dictate them.

  16. Design development of graphite primary structures enables SSTO success

    Science.gov (United States)

    Biagiotti, V. A.; Yahiro, J. S.; Suh, Daniel E.; Hodges, Eric R.; Prior, Donald J.

    1997-01-01

    This paper describes the development of a graphite composite wing and a graphite composite intertank primary structure for application toward Single-Stage to Orbit space vehicles such as those under development in NASA's X-33/Reusable Launch Vehicle (RLV) Program. The trade study and designs are based on a Rockwell vertical take-off and horizontal landing (VTHL) wing-body RLV vehicle. Northrop Grumman's approach using a building block development technique is described. Composite Graphite/Bismaleimide (Gr/BMI) material characterization test results are presented. Unique intertank and wing composite subcomponent test article designs are described and test results to date are presented. Wing and intertank Full Scale Section Test Article (FSTA) objectives and designs are outlined. Trade studies, supporting building block testing, and FSTA demonstrations combine to develop graphite primary structure composite technology that enables developing X-33/RLV design programs to meet critical SSTO structural weight and operations performance criteria.

  17. 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

  18. 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.

  19. Block copolymer based composition and morphology control in nanostructured hybrid materials for energy conversion and storage: solar cells, batteries, and fuel cells

    KAUST Repository

    Orilall, M. Christopher

    2011-01-01

    The development of energy conversion and storage devices is at the forefront of research geared towards a sustainable future. However, there are numerous issues that prevent the widespread use of these technologies including cost, performance and durability. These limitations can be directly related to the materials used. In particular, the design and fabrication of nanostructured hybrid materials is expected to provide breakthroughs for the advancement of these technologies. This tutorial review will highlight block copolymers as an emerging and powerful yet affordable tool to structure-direct such nanomaterials with precise control over structural dimensions, composition and spatial arrangement of materials in composites. After providing an introduction to materials design and current limitations, the review will highlight some of the most recent examples of block copolymer structure-directed nanomaterials for photovoltaics, batteries and fuel cells. In each case insights are provided into the various underlying fundamental chemical, thermodynamic and kinetic formation principles enabling general and relatively inexpensive wet-polymer chemistry methodologies for the efficient creation of multiscale functional materials. Examples include nanostructured ceramics, ceramic-carbon composites, ceramic-carbon-metal composites and metals with morphologies ranging from hexagonally arranged cylinders to three-dimensional bi-continuous cubic networks. The review ends with an outlook towards the synthesis of multicomponent and hierarchical multifunctional hybrid materials with different nano-architectures from self-assembly of higher order blocked macromolecules which may ultimately pave the way for the further development of energy conversion and storage devices. © 2011 The Royal Society of Chemistry.

  20. Block copolymer based composition and morphology control in nanostructured hybrid materials for energy conversion and storage: solar cells, batteries, and fuel cells.

    Science.gov (United States)

    Orilall, M Christopher; Wiesner, Ulrich

    2011-02-01

    The development of energy conversion and storage devices is at the forefront of research geared towards a sustainable future. However, there are numerous issues that prevent the widespread use of these technologies including cost, performance and durability. These limitations can be directly related to the materials used. In particular, the design and fabrication of nanostructured hybrid materials is expected to provide breakthroughs for the advancement of these technologies. This tutorial review will highlight block copolymers as an emerging and powerful yet affordable tool to structure-direct such nanomaterials with precise control over structural dimensions, composition and spatial arrangement of materials in composites. After providing an introduction to materials design and current limitations, the review will highlight some of the most recent examples of block copolymer structure-directed nanomaterials for photovoltaics, batteries and fuel cells. In each case insights are provided into the various underlying fundamental chemical, thermodynamic and kinetic formation principles enabling general and relatively inexpensive wet-polymer chemistry methodologies for the efficient creation of multiscale functional materials. Examples include nanostructured ceramics, ceramic-carbon composites, ceramic-carbon-metal composites and metals with morphologies ranging from hexagonally arranged cylinders to three-dimensional bi-continuous cubic networks. The review ends with an outlook towards the synthesis of multicomponent and hierarchical multifunctional hybrid materials with different nano-architectures from self-assembly of higher order blocked macromolecules which may ultimately pave the way for the further development of energy conversion and storage devices.

  1. Nerve Blocks

    Science.gov (United States)

    ... News Physician Resources Professions Site Index A-Z Nerve Blocks A nerve block is an injection to ... the limitations of Nerve Block? What is a Nerve Block? A nerve block is an anesthetic and/ ...

  2. 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

  3. Investigation of an Alternative Fuel Form for the Liquid Salt Cooled Very High Temperature Reactor (LS-VHTR)

    International Nuclear Information System (INIS)

    Casino, William A. Jr.

    2006-01-01

    Much of the recent studies investigating the use of liquid salts as reactor coolants have utilized a core configuration of graphite prismatic fuel block assemblies with TRISO particles embedded into cylindrical fuel compacts arranged in a triangular pitch lattice. Although many calculations have been performed for this fuel form in gas cooled reactors, it would be instructive to investigate whether an alternative fuel form may yield improved performance for the liquid salt-cooled Very High Temperature Reactor (LS-VHTR). This study investigates how variations in the fuel form will impact the performance of the LS-VHTR during normal and accident conditions and compares the results with a similar analysis that was recently completed for a LS-VHTR core made up of prismatic block fuel. (author)

  4. 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

  5. Status and aspects of fuel element development for advanced high-temperature reactors in the FRG

    International Nuclear Information System (INIS)

    Nickel, H.; Balthesen, E.

    1975-01-01

    In the FRG three basic fuel element designs for application in high temperature gas cooled reactors are being persued: the spherical element, the graphite block element, and the moulded block element (monolith). This report gives the state of development reached with the three types of elements but also views their specific merits and performance margin and presents aspects of their future development potential for operation in advanced HTGR plants. The development of coated feed and breed particles for application in all HTGR fuel elements is treated in more detail. Summarizing it can be said that all the fuel elements as well as their components have proved their aptitude for the dual cycle systems in numerous fuel element and particle performance tests. To adapt these fuel elements and coated particles for advanced reactor concepts and to develop them up to full technical maturity further testing is still necessary, however. Ways of overcoming problems arising from the more stringent requirements are shown. (orig.) [de

  6. Corrosion-induced microstructural changes in a US core graphite

    International Nuclear Information System (INIS)

    Eatherly, W.P.; Lee, D.A.

    1981-01-01

    The results reported here apply to Great Lakes grade H-451 graphite, the core graphite specified for the US HTGR. This graphite is structurally similar to the German reflector grades we have investigated at ORNL, and hence should be applicable to them if similar impurity levels are obtained. Moreover, these results extend and confirm the behavior pattern exhibited by the fuel matrix material A3-3 reported in the previous paper, although the effects are more pronounced in A3-3 presumably due to its resin-type binder and low heat-treatment temperatures

  7. 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.

  8. 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)

  9. Glassy carbon coated graphite for nuclear applications

    International Nuclear Information System (INIS)

    Delpeux S; Cacciaguerra T; Duclaux L

    2005-01-01

    Taking into account the problems caused by the treatment of nuclear wastes, the molten salts breeder reactors are expected to a great development. They use a molten fluorinated salt (mixture of LiF, BeF 2 , ThF 4 , and UF 4 ) as fuel and coolant. The reactor core, made of graphite, is used as a neutrons moderator. Despite of its compatibility with nuclear environment, it appears crucial to improve the stability and inertness of graphite against the diffusion of chemicals species leading to its corrosion. One way is to cover the graphite surface by a protective impermeable deposit made of glassy carbon obtained by the pyrolysis of phenolic resin or polyvinyl chloride precursors. The main difficulty in the synthesis of glassy carbon is to create exclusively, in the primary pyrolysis product, a micro-porosity of about twenty Angstroms which closes later at higher temperature. Therefore, the evacuation of the volatile products occurring mainly between 330 and 600 C, must progress slowly to avoid the material to crack. In this study, the optimal parameters for the synthesis of glassy carbon as well as glassy carbon deposits on nuclear-type graphite pieces are discussed. Both thermal treatment of phenolic and PVC resins have been performed. The structure and micro-texture of glassy carbon have been investigated by X-ray diffraction, scanning and transmission electron microscopies and helium pycno-metry. Glassy carbon samples (obtained at 1200 C) show densities ranging from 1.3 to 1.55 g/cm 3 and closed pores with nano-metric size (∼ 5 to 10 nm) appear clearly on the TEM micrographs. Then, a thermal treatment to 2700 C leads to the shrinkage of the entangled graphene ribbons, in good agreement with the proposed texture model for glassy carbon. Glassy carbon deposits on nuclear graphite have been developed by an impregnation method. The uniformity of the deposit depends clearly on the surface texture and the chemistry of the graphite substrate. The deposit regions where

  10. Glassy carbon coated graphite for nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Delpeux, S.; Cacciaguerra, T.; Duclaux, L. [Orleans Univ., CRMD, CNRS, 45 (France)

    2005-07-01

    Taking into account the problems caused by the treatment of nuclear wastes, the molten salts breeder reactors are expected to a great development. They use a molten fluorinated salt (mixture of LiF, BeF{sub 2}, ThF{sub 4}, and UF{sub 4}) as fuel and coolant. The reactor core, made of graphite, is used as a neutrons moderator. Despite of its compatibility with nuclear environment, it appears crucial to improve the stability and inertness of graphite against the diffusion of chemicals species leading to its corrosion. One way is to cover the graphite surface by a protective impermeable deposit made of glassy carbon obtained by the pyrolysis of phenolic resin [1,2] or polyvinyl chloride [3] precursors. The main difficulty in the synthesis of glassy carbon is to create exclusively, in the primary pyrolysis product, a micro-porosity of about twenty Angstroms which closes later at higher temperature. Therefore, the evacuation of the volatile products occurring mainly between 330 and 600 C, must progress slowly to avoid the material to crack. In this study, the optimal parameters for the synthesis of glassy carbon as well as glassy carbon deposits on nuclear-type graphite pieces are discussed. Both thermal treatment of phenolic and PVC resins have been performed. The structure and micro-texture of glassy carbon have been investigated by X-ray diffraction, scanning and transmission electron microscopies and helium pycno-metry. Glassy carbon samples (obtained at 1200 C) show densities ranging from 1.3 to 1.55 g/cm{sup 3} and closed pores with nano-metric size ({approx} 5 to 10 nm) appear clearly on the TEM micrographs. Then, a thermal treatment to 2700 C leads to the shrinkage of the entangled graphene ribbons (Fig 1), in good agreement with the proposed texture model for glassy carbon (Fig 2) [4]. Glassy carbon deposits on nuclear graphite have been developed by an impregnation method. The uniformity of the deposit depends clearly on the surface texture and the chemistry

  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. Three-Dimensional Analysis of the Hot-Spot Fuel Temperature in Pebble Bed and Prismatic Modular Reactors

    International Nuclear Information System (INIS)

    In, W. K.; Lee, S. W.; Lim, H. S.; Lee, W. J.

    2006-01-01

    High temperature gas-cooled reactors(HTGR) have been reviewed as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor(PBR) and a prismatic modular reactor(PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both a PBR and a PMR. The objective of this study is to predict the hot-spot fuel temperature distributions in a PBR and a PMR at a steady state. The computational fluid dynamics(CFD) code, CFX-10 is used to perform the three-dimensional analysis. The latest design data was used here based on the reference reactor designs, PBMR400 and GTMHR60

  13. Graphite-moderated and heavy water-moderated spectral shift controlled reactors

    International Nuclear Information System (INIS)

    Alcala Ruiz, F.

    1984-01-01

    It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs

  14. Nondestructive testing on graphite structures for high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Kambe, Mamoru; Tsuji, Nobumasa.

    1994-01-01

    The application of ultrasonic (for internal defects) and eddy current testing (for surface defects) were investigated on the structures of nuclear-grade IG-110 and PGX graphite for the HTTR. The equipment were developed in order to detect the specific configuration of graphite blocks and the testing conditions were defined as the practical testing methods. The established testing methods are being used for the acceptance tests of graphite structures in the HTTR. (author)

  15. Radionuclide characterization of graphite stacks from plutonium production reactors of the Siberian group of chemical enterprises

    International Nuclear Information System (INIS)

    Bushuev, A.V.; Verzilov, Yu.M.; Zubarev, V.N.

    2001-01-01

    The residual radionuclide concentrations and distributions in graphite from moderator stack of plutonium production reactors at Tomsk-7 have been investigated. It was found that the dominant activity of graphite is 14 C. To gain information on surface and volume contamination of graphite blocks from the moderator stack, the special sets of samples were collected and assayed. The schemes are proposed for evaluation of individual radionuclide inventories together with results of the evaluations performed. (author)

  16. 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)

  17. Effects of graphite surface roughness on bypass flow computations for an HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Tung, Yu-Hsin, E-mail: touushin@gmail.com [Idaho National Laboratory, P.O. Box 1625, M.S. 3855, Idaho Falls, ID (United States); Johnson, Richard W., E-mail: Rich.Johnson@inl.gov [Idaho National Laboratory, P.O. Box 1625, M.S. 3855, Idaho Falls, ID (United States); Sato, Hiroyuki, E-mail: sato.hiroyuki09@jaea.go.jp [Idaho National Laboratory, P.O. Box 1625, M.S. 3855, Idaho Falls, ID (United States)

    2012-11-15

    Highlights: Black-Right-Pointing-Pointer CFD calculations are made of bypass flow between graphite blocks in HTGR. Black-Right-Pointing-Pointer Several turbulence models are employed to compare to friction and heat transfer correlations. Black-Right-Pointing-Pointer Parameters varied include bypass gap width and surface roughness. Black-Right-Pointing-Pointer Surface roughness causes increases in max fuel and coolant temperatures. Black-Right-Pointing-Pointer Surface roughness does not cause increase in outlet coolant temperature variation. - Abstract: Bypass flow in a prismatic high temperature gas reactor (HTGR) occurs between graphite blocks as they sit side by side in the core. Bypass flow is not intentionally designed to occur in the reactor, but is present because of tolerances in manufacture, imperfect installation and expansion and shrinkage of the blocks from heating and irradiation. It is desired to increase the knowledge of the effects of such flow; it has been suggested that it may be as much as 20% of the total helium coolant flow [INL Report 2007, INL/EXT-07-13289]. Computational fluid dynamic (CFD) simulations can provide estimates of the scale and impacts of bypass flow. Previous CFD calculations have examined the effects of bypass gap width, level and distribution of heat generation and effects of shrinkage. The present contribution examines the effects of graphite surface roughness on the bypass flow for different relative roughness factors for three gap widths. Such calculations should be validated using specific bypass flow measurements. While such experiments are currently underway for the specific reference prismatic HTGR design for the next generation nuclear plant (NGNP) program of the U.S. Dept. of Energy, the data are not yet available. To enhance confidence in the present calculations, wall shear stress and heat transfer results for several turbulence models and their associated wall treatments are first compared for steady flow in a

  18. Inter and Intra Molecular Phase Separation Environment Effects on PI-PEO Block Copolymers for Batteries and Fuel Cells

    Science.gov (United States)

    Xue, Chen-Chen; Meador, Mary Ann B.; Eby, R. K.; Cheng, Stephen Z. D.; Ge, Jason J.; Cubon, Valerie A.

    2002-01-01

    Rod-coil molecules have been introduced as a novel type of block copolymers with unique microstructure due to their ability to self-assemble to various ordered morphologies on a nanometer length scale. These molecules, comprised two homo polymers joined together at one end, microphase separate into ordered, periodic arrays of spheres, cylinders in the bulk state and or solution. To get ordered structure in a reasonable scale, additional force field are applied, such as mechanical shearing, electric field and magnetic field. Recently, progress has made it a possible to develop a new class of polyimides (PI)-Polyethylene oxide (PEO) that are soluble in polar organic solvents. The solvent-soluble PI-PEO has a wide variety of applications in microelectronics, since these PI-PEO films exhibit a high degree of thermal and chemical stability. In this paper, we report the self-assembled ordered structure of PI-PEO molecules formed from concentrate solution.

  19. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    Energy Technology Data Exchange (ETDEWEB)

    Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

    1992-02-01

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.

  20. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    International Nuclear Information System (INIS)

    Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

    1992-02-01

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched 235 U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched 235 U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing

  1. 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

  2. Design Procedure of Graphite Components by ASME HTR Codes

    International Nuclear Information System (INIS)

    Kang, Ji-Ho; Jo, Chang Keun

    2016-01-01

    In this study, the ASME B and PV Code, Subsection HH, Subpart A, design procedure for graphite components of HTRs was reviewed and the differences from metal materials were remarked. The Korean VHTR has a prismatic core which is made of multiple graphite blocks, reflectors, and core supports. One of the design issues is the assessment of the structural integrity of the graphite components because the graphite is brittle and shows quite different behaviors from metals in high temperature environment. The American Society of Mechanical Engineers (ASME) issued the latest edition of the code for the high temperature reactors (HTR) in 2015. In this study, the ASME B and PV Code, Subsection HH, Subpart A, Graphite Materials was reviewed and the special features were remarked. Due the brittleness of graphites, the damage-tolerant design procedures different from the conventional metals were adopted based on semi-probabilistic approaches. The unique additional classification, SRC, is allotted to the graphite components and the full 3-D FEM or equivalent stress analysis method is required. In specific conditions, the oxidation and viscoelasticity analysis of material are required. The fatigue damage rule has not been established yet

  3. Design Procedure of Graphite Components by ASME HTR Codes

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Ji-Ho; Jo, Chang Keun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this study, the ASME B and PV Code, Subsection HH, Subpart A, design procedure for graphite components of HTRs was reviewed and the differences from metal materials were remarked. The Korean VHTR has a prismatic core which is made of multiple graphite blocks, reflectors, and core supports. One of the design issues is the assessment of the structural integrity of the graphite components because the graphite is brittle and shows quite different behaviors from metals in high temperature environment. The American Society of Mechanical Engineers (ASME) issued the latest edition of the code for the high temperature reactors (HTR) in 2015. In this study, the ASME B and PV Code, Subsection HH, Subpart A, Graphite Materials was reviewed and the special features were remarked. Due the brittleness of graphites, the damage-tolerant design procedures different from the conventional metals were adopted based on semi-probabilistic approaches. The unique additional classification, SRC, is allotted to the graphite components and the full 3-D FEM or equivalent stress analysis method is required. In specific conditions, the oxidation and viscoelasticity analysis of material are required. The fatigue damage rule has not been established yet.

  4. 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

  5. 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.

  6. The preliminary feasibility of intercalated graphite railgun armatures

    International Nuclear Information System (INIS)

    Gaier, J.R.; Yashan, D.; Naud, S.

    1991-01-01

    This paper reports on graphite intercalation compounds which may provide an excellent material for the fabrication of electro-magnetic railgun armatures. As a pulse of power is fed into the armature the intercalate could be excited into the plasma state around the edges of the armature, while the bulk of the current would be carried through the graphite block. Such an armature would have desirable characteristics of both diffuse plasma armatures and bulk conduction armatures. In addition, the highly anisotropic nature of these materials could enable the electrical and thermal conductivity to be tailored to meet the specific requirements of electromagnetic railgun armatures. Preliminary investigations have been performed in an attempt to determine the feasibility of using graphite intercalation compounds as railgun armatures. Issues of fabrication, resistivity, stability, and electrical current spreading have been addressed for the case of highly oriented pyrolytic graphite

  7. Measurement of the diffusion length of thermal neutrons inside graphite

    International Nuclear Information System (INIS)

    Ertaud, A.; Beauge, R.; Fauquez, H.; De Laboulay, H.; Mercier, C.; Vautrey, L.

    1948-11-01

    The diffusion length of thermal neutrons inside a given industrial graphite is determined by measuring the neutron density inside a parallelepipedal piling up of graphite bricks (2.10 x 2.10 x 2.442 m). A 3.8 curies (Ra α → Be) source is placed inside the parallelepipedal block of graphite and thin manganese detectors are used. Corrections are added to the unweighted measurements to take into account the effects of the damping of supra-thermal neutrons in the measurement area. These corrections are experimentally deduced from the differential measurements made with a cadmium screen interposed between the source and the first plane of measurement. An error analysis completes the report. The diffusion length obtained is: L = 45.7 cm ± 0.3. The average density of the graphite used is 1.76 and the average apparent density of the piling up is 1.71. (J.S.)

  8. On the Thermal Conductivity Change of Matrix Graphite Materials after Neutron Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Kim, Eung-Seon; Sah, Injin; Park, Daegyu; Kim, Youngjun; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this work, the variations of the thermal conductivity of the A3-3 matrix graphite after neutron irradiation is discussed as well as of the IG-110 graphite for comparison. Neutron irradiation of the graphite specimens was carried out as a part of the first irradiation test of KAERI's coated particle fuel specimens by use of Hanaro research reactor. This work can be summarized as follows: 1) In the evaluation of the specific heat of the graphite materials, various literature data were used and the variations of the specific heat data of all the graphite specimens are observed well agreed, irrespectively of the difference in specimens (graphite and matrix graphite and irradiated and un-irradiated). 2) This implies that it should be reasonable that for both structural graphite and fuel matrix graphite, and even for the neuron-irradiated graphite, any of these specific heat data set be used in the calculation of the thermal conductivity. 3) For the irradiated A3-3 matrix graphite specimens, the thermal conductivity decreased on both directions. On the radial direction, the tendency of variation upon temperature is similar to that of unirradiated specimen, i.e., decreasing as the temperature increases. 4) In the German irradiation experiments with A3-27 matrix graphite specimens, the thermal conductivity of the un-irradiated specimen shows a decrease and that of irradiated specimen is nearly constant as the temperature increases. 5) The thermal conductivity of the irradiated IG-110 was considerably decreased compared with that of un-irradiated specimens The difference of the thermal conductivity of un-irradiated and irradiated IG-110 graphite specimens is much larger than that of un-irradiated and irradiated A3-3 matrix graphite specimens.

  9. Studies on the behavior of graphite structures irradiated in the Dragon Reactor. Dragon Project report

    Energy Technology Data Exchange (ETDEWEB)

    Everett, M. R.; Graham, L. W.; Ridealgh, F.

    1971-11-15

    Design data for the physical and mechanical property changes which occur in graphite structural and fuel body components irradiated in an HTR are largely obtained from small specimens tested in the laboratory and in materials test reactors. A brief data summary is given. This graphite physics data can be used to predict dimensional changes, internal stress generation and strength changes in the graphite materials of HTR fuel elements irradiated in the Dragon Reactor. In this paper, the results which have been obtained from post-irradiation examination of a number of fuel pins, are compared with prediction.

  10. 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)

  11. 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)

  12. Methyl chloride via oxyhydrochlorination of methane: A building block for chemicals and fuels from natural gas. Environmental assessment

    International Nuclear Information System (INIS)

    1996-09-01

    DOE's natural gas mission, in partnership with its stakeholders, is to undertake and promote activities to maximize the Nation's ability to supply, transport, and use natural gas to encourage economic growth, enhance energy interests security, and improve the environment. In implementing this mission, DOE has been involved in promoting domestic natural gas as a clean, abundant, and reliable source of energy. In particular, DOE is interested in technologies capable of converting natural gas to other valuable resources, such as transportation fuels, hydrogen, and premium chemicals. The purpose of the proposed action is to further examine the potential of one such technology for natural gas conversion. Over the past five years, DOE's Pittsburgh Energy Technology Center has supported a research program to determine the feasibility of producing methyl chloride (CH 3 Cl), a key ingredient used in the silicone industry, directly from methane (the primary component of natural gas) via an oxyhydrochlorination (OHC) process. As a result of this research program the OHC process is now ready for further development. The proposed action would advance the OHC natural gas conversion technology to an integrated engineering-scale process at the Dow Corning plant in Carrollton, Kentucky

  13. Improving the chemical compatibility of sealing glass for solid oxide fuel cells: Blocking the reactive species by controlled crystallization

    Science.gov (United States)

    Zhang, Teng; Zou, Qi; Zeng, Fanrong; Wang, Shaorong; Tang, Dian; Yang, Hiswen

    2012-10-01

    The chemical compatibility of sealing glass is of great importance for Solid oxide fuel cell (SOFC). In this work, the interfacial reaction between sealing glass and Cr-containing interconnect alloy is characterized by reacting Cr2O3 powders with a representative SrO-containing glass crystallized by different heat-treatment schedules. The crystalline structure and crystalline content of sealing glass are determined by X-ray diffraction. The results show that the fraction of Cr6+ decreases from 39.8 ± 1.9% for quenched glass to 8.2 ± 0.4% for glass crystallized at 900 °C for 2 h. In addition, the interfacial reaction can be further reduced with increasing crystallization temperature and time as well as the addition of nucleation agent (TiO2). The formation of some Sr-containing crystalline phases, Sr2SiO4 and Sr(TiO3), contributes to the improvement of chemical compatibility of sealing glass, in agreement with the results of thermodynamic calculations.

  14. 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.

  15. 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.

  16. 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

  17. Protection of nuclear graphite toward fluoride molten salt by glassy carbon deposit

    International Nuclear Information System (INIS)

    Bernardet, V.; Gomes, S.; Delpeux, S.; Dubois, M.; Guerin, K.; Avignant, D.; Renaudin, G.; Duclaux, L.

    2009-01-01

    Molten salt reactor represents one of the promising future Generation IV nuclear reactors families where the fuel, a liquid molten fluoride salt, is circulating through the graphite reactor core. The interactions between nuclear graphite and fluoride molten salt and also the graphite surface protection were investigated in this paper by powder X-ray diffraction, micro-Raman spectroscopy and scanning electron microscopy coupled with X-ray microanalysis. Nuclear graphite discs were covered by two kinds of protection deposit: a glassy carbon coating and a double coating of pyrolitic carbon/glassy carbon. Different behaviours have been highlighted according to the presence and the nature of the coated protection film. Intercalation of molten salt between the graphite layers did not occur. Nevertheless the molten salt adhered more or less to the surface of the graphite disc, filled more or less the graphite surface porosity and perturbed more or less the graphite stacking order at the disc surface. The behaviour of unprotected graphite was far to be satisfactory after two days of immersion of graphite in molten salt at 500 deg. C. The best protection of the graphite disc surface, with the maximum of inertness towards molten salt, has been obtained with the double coating of pyrolitic carbon/glassy carbon

  18. 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)

  19. Graphite moderated {sup 252}Cf source

    Energy Technology Data Exchange (ETDEWEB)

    Sajo B, L.; Barros, H.; Greaves, E. D. [Universidad Simon Bolivar, Nuclear Physics Laboratory, Apdo. 89000, 1080A Caracas (Venezuela, Bolivarian Republic of); Vega C, H. R., E-mail: fermineutron@yahoo.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

    2014-08-15

    The thorium molten salt reactor is an attractive and affordable nuclear power option for developing countries with insufficient infrastructure and limited technological capability. In the aim of personnel training and experience gathering at the Universidad Simon Bolivar there is in progress a project of developing a subcritical thorium liquid fuel reactor. The neutron source to run this subcritical reactor is a {sup 252}Cf source and the reactor will use high-purity graphite as moderator. Using the MCNP5 code the neutron spectra of the {sup 252}Cf in the center of the graphite moderator has been estimated along the channel where the liquid thorium salt will be inserted; also the ambient dose equivalent due to the source has been determined around the moderator. (Author)

  20. Process behavior and environmental assessment of 14C releases from an HTGR fuel reprocessing facility

    International Nuclear Information System (INIS)

    Snider, J.W.; Kaye, S.V.

    1976-01-01

    Large quantities of 14 CO 2 will be evolved when graphite fuel blocks are burned during reprocessing of spent fuel from HTGR reactors. The possible release of some or all of this 14 C to the environment is a matter of concern which is investigated in this paper. Various alternatives are considered in this study for decontaminating and releasing the process off-gas to the environment. Concomitant radiological analyses have been done for the waste process scenarios to supply the necessary feedbacks for process design

  1. 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)

  2. Application of quantitative image analysis to the investigation of macroporosity of graphitic materials

    International Nuclear Information System (INIS)

    Delle, W.; Koizlik, K.; Hoven, H.; Wallura, E.

    1978-01-01

    The essence of quantitative image analysis is that the classification of graphitic materials to be inspected is possible on the basis of the grey value contrast between pores (dark) and carbon (bright). Macroporosity is defined as total of all pores with diameters larger than 0.2 μm. The pore size distributions and pore shapes of graphites based on petroleum, pitch, gilsonite and fluid coke as well as graphitic fuel matrices and pyrolytic carbons were investigated. The relationships between maximum grain size, macroporosity and total porosity as well as the anisotropies of macroporosity and electrical resistivity of graphite were established. (orig./GSC) [de

  3. Nonlinear seismic analysis of a graphite reactor core

    International Nuclear Information System (INIS)

    Laframboise, W.L.; Desmond, T.P.

    1988-01-01

    Design and construction of the Department of Energy's N-Reactor located in Richland, Washington was begun in the late 1950s and completed in the early 1960s. Since then, the reactor core's structural integrity has been under review and is considered by some to be a possible safety concern. The reactor core is moderated by graphite. The safety concern stems from the degradation of the graphite due to the effects of long-term irradiation. To assess the safety of the reactor core when subjected to seismic loads, a dynamic time-history structural analysis was performed. The graphite core consists of 89 layers of numerous graphite blocks which are assembled in a 'lincoln-log' lattice. This assembly permits venting of steam in the event of a pressure tube rupture. However, such a design gives rise to a highly nonlinear structure when subjected to earthquake loads. The structural model accounted for the nonlinear interlayer sliding and for the closure and opening of gaps between the graphite blocks. The model was subjected to simulated earthquake loading, and the time-varying response of selected elements critical to safety were monitored. The analytically predicted responses (displacements and strains) were compared to allowable responses to assess margins of safety. (orig.)

  4. Graphite tail powder and liquid biofertilizer as trace elements source for ground nut

    Science.gov (United States)

    Hindersah, Reginawanti; Setiawati, M. Rochimi; Fitriatin, B. Natalie; Suryatama, Pujawati; Asmiran, Priyanka; Panatarani, Camellia; Joni, I. Made

    2018-02-01

    Utilization of graphite tail waste from the mineral beneficiation processing is very important since it contain significant amount of essential minerals which are necessary for plant growth. These mineral are required in biochemical processes and mainly play an important role as cofactor in enzymatic reaction. The objective of this research is to investigate the performance of graphite tail on supporting plant growth and yield of ground nut (Arachishypogeae L.). A field experiment has been performed to test the performance of mixed graphite tail and reduced organic matter dose. The graphite tail size were reduced to various sieved size, -80 mesh, -100 mesh and -200 mesh. The experiment was setup in randomized block design with 4 treatments and 6 replications for each treatment, while the control plot is received without graphite tail. The results demonstrated that reduced organic matter along with -200 mesh tail has potentially decreased plant height at the end of vegetative growth stage, in contrast for to -80 mesh tail amendment increased individual fresh plant biomass. Statistically, there was no change of plant nodule, individual shoot fresh and dry weight, root nodule, number of pod following any mesh of graphite tail amendment. Reducing organic matter while adding graphite tail of 5% did not change bean weight in all plot. In contrast, reduced organic matter along with 80-mesh graphite tail amendment improved the nut yield per plot. This experiment suggests that graphite tail, mainly -80 mesh graphite tail can be possibly used in legume production.

  5. Gamma scanning of full scale HTR fuel elements

    International Nuclear Information System (INIS)

    Harrison, T.A.; Simpson, J.A.H.; Nabielek, H.

    1983-04-01

    Gamma scanning for the determination of burn-up and fission product inventory has been developed at the Dragon Project, suitable for measurements on fuel elements and segments from full-sized integral block elements. This involved the design and construction of a new lead flask with sophisticated collimator design. State-of-the art gamma spectrometric equipment was set up to cope with strong variations of count-rate and high data throughput. Software efforts concentrated on the calculation of the self absorption and absorption corrections in the complicated geometry of multi-hole graphite block segments with a corrugated circumference. The techniques described here are applicable to the non-destructive examination of a wide range of fuel element designs. (author)

  6. Qualitative and quantitative analysis of N-acetyllactosamine and lacto-N-biose, the two major building blocks of human milk oligosaccharides in human milk samples by high-performance liquid chromatography-tandem mass spectrometry using a porous graphitic carbon column.

    Science.gov (United States)

    Balogh, Réka; Jankovics, Péter; Béni, Szabolcs

    2015-11-27

    This study presents a validated, porous graphitic carbon stationary phase-based LC-MS/MS method for the identification and quantification of lacto-N-biose (LNB) and N-acetyllactosamine (LacNAc). These compounds are the major building blocks of human milk oligosaccharides, however the presence of their unbound form in human milk has not been examined so far. The separation of these highly related structures in their alditol form was accomplished by a gradient LC method and multiple reaction monitoring (MRM) analysis after appropriate sample preparation including size-exclusion chromatography and solid-phase extraction. Baseline separation of the components provides the selectivity for the method. Validation was performed according to the European Medicines Agency (EMA) Guidelines and the method was found to be precise and accurate. Using our developed and validated method we were able to identify and quantify both saccharides in human milk for the first time. Based on our results the LacNAc concentration is in the range of 6.7-31μg/mL while LNB concentration decreased from 26μg/mL below the detection limit during the first week of lactation. The presence of LNB and LacNAc in human milk also implies new biological functions which can lead us closer to the understanding of the various functions of this complex biofluid. Copyright © 2015 Elsevier B.V. All rights reserved.

  7. 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

  8. 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

  9. Management of radioactive waste in nuclear power: handling of irradiated graphite from water-cooled graphite reactors

    International Nuclear Information System (INIS)

    Anfimov, S.S.

    2000-01-01

    As a result of decommissioning of water-cooled graphite-moderated reactors, a large amount of rad-waste in the form of graphite stack fragments is generated (on average 1500-2000 tons per reactor). That is why it is essentially important, although complex from the technical point of view, to develop advanced technologies based on up-to-date remotely-controlled systems for unmanned dismantling of the graphite stack containing highly-active long-lived radionuclides and for conditioning of irradiated graphite (IG) for the purposes of transportation and subsequent long term and ecologically safe storage either on NPP sites or in special-purpose geological repositories. The main characteristics critical for radiation and nuclear hazards of the graphite stack are as follows: the graphite stack is contaminated with nuclear fuel that has gotten there as a result of the accidents; the graphite mass is 992 tons, total activity -6?104 Ci (at the time of unit shutdown); the fuel mass in the reactor stack amounts to 100-140 kg, as estimated by IPPE and RDIPE, respectively; γ-radiation dose rate in the stack cells varies from 4 to 4300 R/h, with the prevailing values being in the range from 50 to 100 R/h. In this paper the traditional methods of rad-waste handling as bituminization technology, cementing technology are discussed. In terms of IG handling technology two lines were identified: long-term storage of conditioned IG and IG disposal by means of incineration. The specific cost of graphite immobilization in a radiation-resistant polymeric matrix amounts to -2600 USD per 1 t of graphite, whereas the specific cost of immobilization in slag-stone containers with an inorganic binder (cement) is -1400 USD per 1 t of graphite. On the other hand, volume of conditioned IG rad-waste subject for disposal, if obtained by means of the first technology, is 2-2.5 times less than the volume of rad-waste generated by means of the second technology. It can be concluded from the above that

  10. Bypass Flow and Hot Spot Analysis for PMR200 Block-Core Design with Core Restraint Mechanism

    International Nuclear Information System (INIS)

    Lim, Hong Sik; Kim, Min Hwan

    2009-01-01

    The accurate prediction of local hot spot during normal operation is important to ensure core thermal margin in a very high temperature gas-cooled reactor because of production of its high temperature output. The active cooling of the reactor core determining local hot spot is strongly affected by core bypass flows through the inter-column gaps between graphite blocks and the cross gaps between two stacked fuel blocks. The bypass gap sizes vary during core life cycle by the thermal expansion at the elevated temperature and the shrinkage/swelling by fast neutron irradiation. This study is to investigate the impacts of the variation of bypass gaps during core life cycle as well as core restraint mechanism on the amount of bypass flow and thus maximum fuel temperature. The core thermo fluid analysis is performed using the GAMMA+ code for the PMR200 block-core design. For the analysis not only are some modeling features, developed for solid conduction and bypass flow, are implemented into the GAMMA+ code but also non-uniform bypass gap distribution taken from a tool calculating the thermal expansion and the shrinkage/swell of graphite during core life cycle under the design options with and without core restraint mechanism is used

  11. 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.

  12. 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.

  13. Graphite-moderated and heavy water-moderated spectral shift controlled reactors; Reactores de moderador solido controlados por desplazamiento espectral

    Energy Technology Data Exchange (ETDEWEB)

    Alcala Ruiz, F

    1984-07-01

    It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs.

  14. 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.

  15. 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.

  16. On the Crystallization of Compacted and Chunky Graphite from Liquid Multicomponent Iron-Carbon-Silicon-Based Melts

    Science.gov (United States)

    Stefanescu, D. M.; Huff, R.; Alonso, G.; Larrañaga, P.; De la Fuente, E.; Suarez, R.

    2016-08-01

    Extensive SEM work was carried out on deep-etched specimens to reveal the evolution of compacted and chunky graphite in magnesium-modified multicomponent Fe-C-Si alloys during early solidification and at room temperature. The findings of this research were then integrated in the current body of knowledge to produce an understanding of the crystallization of compacted and chunky graphite. It was confirmed that growth from the liquid for both compacted and chunky graphite occurs radially from a nucleus, as foliated crystals and dendrites. The basic building blocks of the graphite aggregates are hexagonal faceted graphite platelets with nanometer height and micrometer width. Thickening of the platelets occurs through growth of additional graphene layers nucleated at the ledges of the graphite prism. Additional thickening resulting in complete joining of the platelets may occur from the recrystallization of the amorphous carbon that has diffused from the liquid through the austenite, once the graphite aggregate is enveloped in austenite. With increasing magnesium levels, the foliated graphite platelets progressively aggregate along the c-axis forming clusters. The clusters that have random orientation, eventually produce blocky graphite, as the spaces between the parallel platelets disappear. This is typical for compacted graphite irons and tadpole graphite. The chunky graphite aggregates investigated are conical sectors of graphite platelets stacked along the c-axis. The foliated dendrites that originally develop radially from a common nucleus may aggregate along the c-axis forming blocky graphite that sometimes exhibits helical growth. The large number of defects (cavities) observed in all graphite aggregates supports the mechanism of graphite growth as foliated crystals and dendrites.

  17. Effect of a Central Graphite Column on a Pebble Flow in a Pebble Bed Core

    International Nuclear Information System (INIS)

    In, W. K.; Lee, W. J.; Chang, J. H.

    2006-01-01

    A pebble bed reactor(PBR) uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. The pebble bed core is configured as cylindrical or annular depending on the reactor power. It is well known that an annular core can increase a cores' thermal power. The annular inner core zone is typically filled with movable graphite balls or a fixed graphite column. The first problem with this conventional annular core is that it is difficult to maintain a boundary between the central graphite ball zone and the outer fuel zone. The second problem is that it is expensive to replace the central fixed graphite column after several tens of years of reactor operation. In order to resolve these problems, a PBR with a central graphite column in a low core is invented. This paper presents the effect of the central graphite column on a pebble flow by using the computational fluid dynamics(CFD) code, CFX-10

  18. 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

  19. 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.

  20. On possibility of degradation of lava-like fuel-containing materials of the 4-th block of Chernobyl NPP under internal self-irradiation by alpha-particle sources

    International Nuclear Information System (INIS)

    Pazukhin, Eh. M.; Borovoj, A.A.; Rudya, K.G.

    2002-01-01

    It is shown that internal self-irradiation by alpha-particle beam cannot be a cause of change of strength characteristics of silicate matrix and so a cause of degradation of Chernobyl lava-like materials. A new method is proposed for management with lava-like fuel-containing materials of the 4-th block: vitrification in smelter unit situated in bubbler-basin and storage of prepared immobilized compacts in corresponding depositories [ru

  1. 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.

  2. High-performance supercapacitors based on hierarchically porous graphite particles

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Zheng; Wen, Jing; Yan, Chunzhu; Rice, Lynn; Sohn, Hiesang; Lu, Yunfeng [Department of Chemical and Biomolecular Engineering, University of California, Los Angeles, CA 90095 (United States); Shen, Meiqing [School of Chemical Engineering and Technology, Tianjin University, Tianjin, 300072 (China); Cai, Mei [General Motor R and D Center, Warren, MI 48090 (United States); Dunn, Bruce [Department of Materials Science and Engineering, University of California, Los Angeles, CA 90095 (United States)

    2011-07-15

    Hierarchically porous graphite particles are synthesized using a continuous, scalable aerosol approach. The unique porous graphite architecture provides the particles with high surface area, fast ion transportation, and good electronic conductivity, which endows the resulting supercapacitors with high energy and power densities. This work provides a new material platform for high-performance supercapacitors with high packing density, and is adaptable to battery electrodes, fuel-cell catalyst supports, and other applications. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  3. Epidural block

    Science.gov (United States)

    ... page: //medlineplus.gov/ency/patientinstructions/000484.htm Epidural block - pregnancy To use the sharing features on this page, please enable JavaScript. An epidural block is a numbing medicine given by injection (shot) ...

  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. Studies on the graphite rupture under irradiation induced strains

    International Nuclear Information System (INIS)

    Jouquet, G.; Berthion, Y.; L'Homme, A.

    1980-01-01

    Following the RMG experiments (failure of graphite by mechanical effect, i.e. under very high temperature gradient) an experimental program called RWG (Failure of Graphite by WIGNER effect) was initiated in 75 at C.E.A. 3 experiments have been already performed in the OSIRIS reactor at Saclay: RWG 01, 02 and 03. A 4th one, RWG04, is scheduled for the end of 79, may be in collaboration with GERMANY. The aim of the RWG experiments is to induce internal stresses in graphite blocks by irradiation at high temperature which would lead or not to their failure so one could bracket, as tightly as possible, the critical value for failure onset in given experimental conditions

  6. Characterization, treatment and conditioning of radioactive graphite from decommissioning of nuclear reactors

    International Nuclear Information System (INIS)

    2006-09-01

    Graphite has been used as a moderator and reflector of neutrons in more than 100 nuclear power plants and in many research and plutonium-production reactors. It is used primarily as a neutron reflector or neutron moderator, although graphite is also used for other features of reactor cores, such as fuel sleeves. Many of the graphite-moderated reactors are now quite old, with some already shutdown. Therefore radioactive graphite dismantling and the management of radioactive graphite waste are becoming an increasingly important issue for a number of IAEA Member States. Worldwide, there are more than 230 000 tonnes of radioactive graphite which will eventually need to be managed as radioactive waste. Proper management of radioactive graphite waste requires complex planning and the implementation of several interrelated operations. There are two basic options for graphite waste management: (1) packaging of non-conditioned graphite waste with subsequent direct disposal of the waste packages, and (2) conditioning of graphite waste (principally either by incineration or calcination) with separate disposal of any waste products produced, such as incinerator ash. In both cases, the specific properties of graphite - such as Wigner energy, graphite dust explosibility, and radioactive gases released from waste graphite - have a potential impact on the safety of radioactive graphite waste management and need to be carefully considered. Radioactive graphite waste management is not specifically addressed in IAEA publications. Only general and limited information is available in publications dealing with decommissioning of nuclear reactors. This report provides a comprehensive discussion of radioactive graphite waste characterization, handling, conditioning and disposal throughout the operating and decommissioning life cycle. The first draft report was prepared at a meeting on 23-27 February 1998. A technical meeting (TM) was held in October 1999 in coincidence with the Seminar on

  7. 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...

  8. Population Blocks.

    Science.gov (United States)

    Smith, Martin H.

    1992-01-01

    Describes an educational game called "Population Blocks" that is designed to illustrate the concept of exponential growth of the human population and some potential effects of overpopulation. The game material consists of wooden blocks; 18 blocks are painted green (representing land), 7 are painted blue (representing water); and the remaining…

  9. Reactor core with rod-shaped fuel cells

    International Nuclear Information System (INIS)

    Dworak, A.

    1977-01-01

    The aim is an optimization of load distribution in the core so that the load decreases in the direction of coolant flow (with gas cooling from above downwards) but so that it remains constant in horizontal layers to the edge of the core. The former produces optimum cooling, because the coolant has to take up decreasing heat output in the direction of flow. The latter simplifies refueling, because replacement of a whole layer having the same burn-up takes place. The upper two layers with the highest output and the shortest dwell time are replaced every 300 days, for example, the third layer is replaced after double this time and 5 more layers after four times this dwell time. After the simultaneous replacement of all layers, the reactor is in the same state as at commissioning. The fuel cells consist of hexagonal graphite blocks about 1.65 metres in height and 0.75 wide, for example. Each block contains about 100 through cooling channels and about 200 fuel channels closed on both sides. A large number of columns each consisting of 8 blocks is arranged in a tight honeycomb pattern and forms the core. Within each of the 8 horizontal layers of blocks, each fuel cell contains the same fuel mixture with predetermined dwell time. The fuel mixture is suited to the dwell time planned for each layer. The various fuel cells are kept at the same output by burnable neutron poisons in special channels provided for this purpose in the fuel cell and/or by absorber rods, or a planned load distribution is maintained. (HP) [de

  10. Destruction of nuclear graphite using closed chamber incineration

    International Nuclear Information System (INIS)

    Senor, D.J.; Hollenberg, G.W.; Morgan, W.C.; Marianowski, L.G.

    1994-01-01

    Closed chamber incineration (CCI) is a novel technique by which irradiated nuclear graphite may be destroyed without the risk of radioactive cation release into the environment. The process utilizes an enclosed combustion chamber coupled with molten carbonate fuel cells (MCFCs). The transport of cations is intrinsically suppressed by the MCFCs, such that only the combustion gases are conducted through for release to the environment. An example CCI design was developed which had as its goal the destruction of graphite fuel elements from the Fort St. Vrain reactor (FSVR). By employing CCI, the volume of high level waste from the FSVR will be reduced by approximately 87 percent. Additionally, the incineration process will convert the SiC coating on the FSVR fuel particles to SiO 2 , thus creating a form potentially suitable for direct incorporation in a vitrification process stream. The design is compact, efficient, and makes use of currently available technology

  11. 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

  12. 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)

  13. 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.

  14. 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

  15. Inert annealing of irradiated graphite by inductive heating

    International Nuclear Information System (INIS)

    Botzem, W.; Woerner, J.

    2001-01-01

    Fission neutrons change physical properties of graphite being used in nuclear reactors as moderator and as structural material. The understanding of these effects on an atomic model is expressed by dislocations of carbon atoms within the graphite and the thereby stored energy is known as Wigner Energy. The dismantling of the Pile 1 core may necessitate the thermal treatment of the irradiated but otherwise undamaged graphite. This heat treatment - usually called annealing - initiates the release of stored Wigner Energy in a controlled manner. This energy could otherwise give rise to an increase in temperature under certain conditions during transport or preparation for final storage. In order to prevent such an effect it is intended to anneal the major part of Pile 1 graphite before it is packed into boxes suitable for final disposal. Different heating techniques have been assessed. Inductive heating in an inert atmosphere was selected for installation in the Pile 1 Waste Processing Facility built for the treatment and packaging of the dismantled Pile 1 waste. The graphite blocks will be heated up to 250 deg. C in the annealing ovens, which results in the release of significant amount of the stored energy. External heat sources in a final repository will never heat up the storage boxes to such a temperature. (author)

  16. Oxidation of graphites for core support post in air at high temperatures

    International Nuclear Information System (INIS)

    Imai, Hisashi; Fujii, Kimio; Kurosawa, Takeshi

    1982-07-01

    Oxidation reactions of candidate graphites for core support post with atmospheric air were studied in a temperature range between 550 0 C and 1000 0 C. The reaction rates, temperature dependence of the rates and distribution of bulk density in the oxidized graphites were measured and the characters obtained were compared between the brand of graphites. On the basis of the experimental results, dimension and strength of the post after corrosion with air, which might be introduced in rupture accident of primary coolant tube, were discussed. In the case of IG-11 graphite, it was proved that the strength of post is still sufficient even 100 hours after the beginning of the accident and that, however, it is necessary to insert more deeply the post against graphite blocks. (author)

  17. Transport of fission products in matrix and graphite

    International Nuclear Information System (INIS)

    Hoinkis, E.

    1983-06-01

    In the past years new experimental methods were applied to or developed for the investigation of fission product transport in graphitic materials and to characterization of the materials. Models for fission product transport and computer codes for the calculation of core release rates were improved. Many data became available from analysis of concentration profiles in HTR-fuel elements. New work on the effect on diffusion of graphite corrosion, fast neutron flux and fluence, heat treatment, chemical interactions and helium pressure was reported on recently or was in progress in several laboratories. It seemed to be the right time to discuss the status of transport of metallic fission products in general, and in particular the relationship between structural and transport properties. Following a suggestion a Colloquium was organized at the HMI Berlin. Interdisciplinary discussions were stimulated by only inviting a limited number of participants who work in different fields of graphite and fission product transport research. (orig./RW)

  18. 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)

  19. Compacted graphite iron: Cast iron makes a comeback

    Science.gov (United States)

    Dawson, S.

    1994-08-01

    Although compacted graphite iron has been known for more than four decades, the absence of a reliable mass-production technique has resulted in relatively little effort to exploit its operational benefits. However, a proven on-line process control technology developed by SinterCast allows for series production of complex components in high-quality CGI. The improved mechanical properties of compacted graphite iron relative to conventional gray iron allow for substantial weight reduction in gasoline and diesel engines or substantial increases in horsepower, or an optimal combination of both. Concurrent with these primary benefits, CGI also provides significant emissions and fuel efficiency benefits allowing automakers to meet legislated performance standards. The operational and environmental benefits of compacted graphite iron together with its low cost and recyclability reinforce cast iron as a prime engineering material for the future.

  20. One new route to optimize the oxidation resistance of TiC/hastelloy (Ni-based alloy) composites applied for intermediate temperature solid oxide fuel cell interconnect by increasing graphite particle size

    Science.gov (United States)

    Qi, Qian; Liu, Yan; Wang, Lujie; Zhang, Hui; Huang, Jian; Huang, Zhengren

    2017-09-01

    TiC/hastelloy composites with suitable thermal expansion and excellent electrical conductivity are promising candidates for IT-SOFC interconnect. In this paper, the TiC/hastelloy composites are fabricated by in-situ reactive infiltration, and the oxidation resistance of composites is optimized by increasing graphite particle size. Results show that the increase of graphite particles size from 1 μm to 40 μm reduces TiC particle size from 2.68 μm to 2.22 μm by affecting the formation process of TiC. Moreover, the decrease of TiC particles size accelerates the fast formation of dense and continuous TiO2/Cr2O3 oxide layer, which bring down the mass gain (800 °C/100 h) from 2.03 mg cm-2 to 1.18 mg cm-2. Meanwhile, the coefficient of thermal expansion decreases from 11.15 × 10-6 °C-1 to 10.80 × 10-6 °C-1, and electrical conductivity maintains about 5800 S cm-1 at 800 °C. Therefore, the decrease of graphite particle size is one simple and effective route to optimize the oxidation resistance of composites, and meantime keeps suitable thermal expansion and good electrical conductivity.

  1. 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.

  2. 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.

  3. Nuclear fuel element

    International Nuclear Information System (INIS)

    Knowles, A.N.

    1979-01-01

    A nuclear fuel-containing body for a high temperature gas cooled nuclear reactor is described which comprises a flat plate in which the nuclear fuel is contained as a dispersion of fission product-retaining coated fuel particles in a flat sheet of graphitic or carbonaceous matrix material. The flat sheet is clad with a relatively thin layer of unfuelled graphite bonded to the sheet by being formed initially from a number of separate preformed graphitic artefacts and then platen-pressed on to the exterior surfaces of the flat sheet, both the matrix material and the artefacts being in a green state, to enclose the sheet. A number of such flat plates are supported edge-on to the coolant flow in the bore of a tube made of neutron moderating material. Where a number of tiers of plates are superimposed on one another, the abutting edges are chamfered to reduce vibration. (author)

  4. 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)

  5. 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

  6. 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)

  7. 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

  8. 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

  9. Study on erbium loading method to improve reactivity coefficients for low radiotoxic spent fuel HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Fukaya, Y., E-mail: fukaya.yuji@jaea.go.jp; Goto, M.; Nishihara, T.

    2015-11-15

    Highlights: • We attempted and optimized erbium loading methods to improve reactivity coefficients for LRSF-HTGR. • We elucidated the mechanism of the improvements for each erbium loading method by using the Bondarenko approach. • We concluded the erbium loading method by embedding into graphite shaft is preferable. - Abstract: Erbium loading methods are investigated to improve reactivity coefficients of Low Radiotoxic Spent Fuel High Temperature Gas-cooled Reactor (LRSF-HTGR). Highly enriched uranium is used for fuel to reduce the generation of toxicity from uranium-238. The power coefficients are positive without the use of any additive. Then, the erbium is loaded into the core to obtain negative reactivity coefficients owing to the large resonance the peak of neutron capture reaction of erbium-167. The loading methods are attempted to find the suitable method for LRSF-HTGR. The erbium is mixed in a CPF fuel kernel, loaded by binary packing with fuel particles and erbium particles, and embedded into the graphite shaft deployed in the center of the fuel compact. It is found that erbium loading causes negative reactivity as moderator temperature reactivity, and from the viewpoint of heat transfer, it should be loaded into fuel pin elements for pin-in-block type fuel. Moreover, the erbium should be incinerated slowly to obtain negative reactivity coefficients even at the End Of Cycle (EOC). A loading method that effectively causes self-shielding should be selected to avoid incineration with burn-up. The incineration mechanism is elucidated using the Bondarenko approach. As a result, it is concluded that erbium embedded into graphite shaft is preferable for LRSF-HTGR to ensure that the reactivity coefficients remain negative at EOC.

  10. 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)

  11. 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...

  12. 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...

  13. Measurement of the diffusion length of thermal neutrons inside graphite; Mesure de la longueur de diffusion des neutrons thermiques dans le graphite

    Energy Technology Data Exchange (ETDEWEB)

    Ertaud, A; Beauge, R; Fauquez, H; De Laboulay, H; Mercier, C; Vautrey, L

    1948-11-01

    The diffusion length of thermal neutrons inside a given industrial graphite is determined by measuring the neutron density inside a parallelepipedal piling up of graphite bricks (2.10 x 2.10 x 2.442 m). A 3.8 curies (Ra {alpha} {yields} Be) source is placed inside the parallelepipedal block of graphite and thin manganese detectors are used. Corrections are added to the unweighted measurements to take into account the effects of the damping of supra-thermal neutrons in the measurement area. These corrections are experimentally deduced from the differential measurements made with a cadmium screen interposed between the source and the first plane of measurement. An error analysis completes the report. The diffusion length obtained is: L = 45.7 cm {+-} 0.3. The average density of the graphite used is 1.76 and the average apparent density of the piling up is 1.71. (J.S.)

  14. 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

  15. 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)

  16. 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)

  17. 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)

  18. A German research project about applicable graphite cutting techniques

    International Nuclear Information System (INIS)

    Holland, D.; Quade, U.; Bach, F.W.; Wilk, P.

    2001-01-01

    In Germany, too, quite large quantities of irradiated nuclear graphite, used in research and prototype reactors, are waiting for an environmental way of disposal. While incineration of nuclear graphite does not seem to be a publicly acceptable way, cutting and packaging into ductile cast iron containers could be a suitable way of disposal in Germany. Nevertheless, the cutting of graphite is also a very difficult technique by which a large amount of secondary waste or dust might occur. An applicable graphite cutting technique is needed. Therefore, a group of 13 German partners, consisting of one university, six research reactor operators, one technical inspection authority, three engineering companies, one industrial cutting specialist and one commercial dismantling company, decided in 1999 to start a research project to develop an applicable technique for cutting irradiated nuclear graphite. Aim of the project is to find the most suitable cutting techniques for the existing shapes of graphite blocks with a minimum of waste production rate. At the same time it will be learned how to sample the dust and collect it in a filter system. The following techniques will be tested and evaluated: thermal cutting, water jet cutting, mechanical cutting with a saw, plasma arc cutting, drilling. The subsequent evaluation will concentrate on dust production, possible irradiation of staff, time and practicability under different constraints. This research project is funded by the German Minister of Education and Research under the number 02 S 7849 for a period of two years. A brief overview about the work to be carried out in the project will be given. (author)

  19. Method of making nuclear fuel bodies

    International Nuclear Information System (INIS)

    Davis, D.E.; Leary, D.F.

    1977-01-01

    A method of making nuclear fuel bodies is described comprising: providing particulate graphite having a particle size not greater than about 1500 microns; impregnating the graphite with a polymerizable organic resin in liquid form; treating the impregnated particles with a hot aqueous acid solution to pre-cure the impregnated resin and to remove excess resin from the surfaces of said graphite particles; heating the treated particles to polymerize the impregnant; blending the impregnated particles with particulate nuclear fuel; and forming a nuclear fuel body by joining the blend of particles into a cohesive mass using a carbonaceous binder

  20. 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

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

    International Nuclear Information System (INIS)

    Radulescu, H.

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    H. radulescu

    2001-09-28

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

  3. QUARTERLY PROGRESS REPORT JANUARY, FEBRUARY, MARCH, 1968 REACTOR FUELS AND MATERIALS DEVELOPMENT PROGRAMS FOR FUELS AND MATERIALS BRANCH OF USAEC DIVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Cadwell, J. J.; de Halas, D. R.; Nightingale, R. E.; Worlton, D. C.

    1968-06-01

    Progress is reported in these areas: nuclear graphite; fuel development for gas-cooled reactors; HTGR graphite studies; nuclear ceramics; fast-reactor nitrides research; non-destructive testing; metallic fuels; basic swelling studies; ATR gas and water loop operation and maintenance; reactor fuels and materials; fast reactor dosimetry and damage analysis; and irradiation damage to reactor metals.

  4. Optimization of the fuel cell of a spectral shift controlled reactor

    International Nuclear Information System (INIS)

    Alcala, F.

    1984-01-01

    Some low enriched uranium-graphite watercooled cells are analyzed from the point of view of both neutronic and thermal behavior. Such cells are of greater interest if their reactivity control is carried out by means of spectral shifting. This may be achieved by changing the relative concentration of a mixture of heavy and light water used as a coolant that flows through a system of tubes arranged in the graphite blocks. The increase of the relative fraction of light water during the burnup cycle extends the length of the cycle and makes it less proliferative. Performance of a cell of this kind is compared with that of a typical pressurized water reactor cell having the same type of fuel, degree of enrichment, specific power, and total power output

  5. The statistical analysis techniques to support the NGNP fuel performance experiments

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh T., E-mail: Binh.Pham@inl.gov; Einerson, Jeffrey J.

    2013-10-15

    This paper describes the development and application of statistical analysis techniques to support the Advanced Gas Reactor (AGR) experimental program on Next Generation Nuclear Plant (NGNP) fuel performance. The experiments conducted in the Idaho National Laboratory’s Advanced Test Reactor employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule. The tests are instrumented with thermocouples embedded in graphite blocks and the target quantity (fuel temperature) is regulated by the He–Ne gas mixture that fills the gap volume. Three techniques for statistical analysis, namely control charting, correlation analysis, and regression analysis, are implemented in the NGNP Data Management and Analysis System for automated processing and qualification of the AGR measured data. The neutronic and thermal code simulation results are used for comparative scrutiny. The ultimate objective of this work includes (a) a multi-faceted system for data monitoring and data accuracy testing, (b) identification of possible modes of diagnostics deterioration and changes in experimental conditions, (c) qualification of data for use in code validation, and (d) identification and use of data trends to support effective control of test conditions with respect to the test target. Analysis results and examples given in the paper show the three statistical analysis techniques providing a complementary capability to warn of thermocouple failures. It also suggests that the regression analysis models relating calculated fuel temperatures and thermocouple readings can enable online regulation of experimental parameters (i.e. gas mixture content), to effectively maintain the fuel temperature within a given range.

  6. A Statistical Analysis on the Coating Layer Thicknesses of a TRISO of 350 MWth Block-type HTR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Min; Jo, C. K.; Cho, M. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    A tri-isotropic coated fuel particle (TRISO) is a basic fuel element of a high temperature reactor (HTR). The block-type HTR fuel is a cylindrical graphite compact in which a large number of TRISOs are embedded. There are more than 11 billion TRISOs in a 350 MW{sub th} block-type HTR core. Among the RSM quadratic models, the BBD model produces the smallest errors at both interior and exterior points. The errors in the quadratic model of the small-type CCD is the biggest, particularly at exterior points. The CCD has a disadvantage of generating a number of decimal places in its factor levels because of its axial points. It is recommended to use the BBD or the full-type CCD with an adjusted axial point which does not produce the decimal places in its factor levels. More general statistical model for a TRISO design will be secured when the number of factors and responses increases. This study treats a statistical analysis on the optimal layer thicknesses of a UCO TRISO of 350 MW{sub th} block-type HTR which cause a minimum tangential stress to act on the SiC layer. Three response surface methods (RSMs) are used as statistical methods and their resulting quadratic models are compared.

  7. A Statistical Analysis on the Coating Layer Thicknesses of a TRISO of 350 MWth Block-type HTR

    International Nuclear Information System (INIS)

    Kim, Young Min; Jo, C. K.; Cho, M. S.

    2016-01-01

    A tri-isotropic coated fuel particle (TRISO) is a basic fuel element of a high temperature reactor (HTR). The block-type HTR fuel is a cylindrical graphite compact in which a large number of TRISOs are embedded. There are more than 11 billion TRISOs in a 350 MW_t_h block-type HTR core. Among the RSM quadratic models, the BBD model produces the smallest errors at both interior and exterior points. The errors in the quadratic model of the small-type CCD is the biggest, particularly at exterior points. The CCD has a disadvantage of generating a number of decimal places in its factor levels because of its axial points. It is recommended to use the BBD or the full-type CCD with an adjusted axial point which does not produce the decimal places in its factor levels. More general statistical model for a TRISO design will be secured when the number of factors and responses increases. This study treats a statistical analysis on the optimal layer thicknesses of a UCO TRISO of 350 MW_t_h block-type HTR which cause a minimum tangential stress to act on the SiC layer. Three response surface methods (RSMs) are used as statistical methods and their resulting quadratic models are compared

  8. Cooling device for thermonuclear reactor and modular packing block for the wall realization of a such device

    International Nuclear Information System (INIS)

    Archer, J.; Stalport, G.; Besson, D.; Faron, R.; Coulon, M.

    1988-01-01

    The cooling device for a thermonuclear reactor wall is made by modular thermally conductive heat-resistant blocks (graphite by example), a prismatic head on one face of each block, the opposite face bearing against cooling tubes, a base to each block with an aperture and rods passing through the apertures reversibly fixing each row of blocks to a support [fr

  9. Chemical analysis of high purity graphite

    International Nuclear Information System (INIS)

    1993-03-01

    The Sub-Committee on Chemical Analysis of Graphite was organized in April 1989, under the Committee on Chemical Analysis of Nuclear Fuels and Reactor Materials, JAERI. The Sub-Committee carried out collaborative analyses among eleven participating laboratories for the certification of the Certified Reference Materials (CRMs), JAERI-G5 and G6, after developing and evaluating analytical methods during the period of September 1989 to March 1992. The certified values were given for ash, boron and silicon in the CRM based on the collaborative analysis. The values for ten elements (Al, Ca, Cr, Fe, Mg, Mo, Ni, Sr, Ti, V) were not certified, but given for information. Preparation, homogeneity testing and chemical analyses for certification of reference materials were described in this paper. (author) 52 refs

  10. High-Temperature Graphite/Phenolic Composite

    Science.gov (United States)

    Seal, Ellis C.; Bodepudi, Venu P.; Biggs, Robert W., Jr.; Cranston, John A.

    1995-01-01

    Graphite-fiber/phenolic-resin composite material retains relatively high strength and modulus of elasticity at temperatures as high as 1,000 degrees F. Costs only 5 to 20 percent as much as refractory materials. Fabrication composite includes curing process in which application of full autoclave pressure delayed until after phenolic resin gels. Curing process allows moisture to escape, so when composite subsequently heated in service, much less expansion of absorbed moisture and much less tendency toward delamination. Developed for nose cone of external fuel tank of Space Shuttle. Other potential aerospace applications for material include leading edges, parts of nozzles, parts of aircraft engines, and heat shields. Terrestrial and aerospace applications include structural firewalls and secondary structures in aircraft, spacecraft, and ships. Modified curing process adapted to composites of phenolic with other fiber reinforcements like glass or quartz. Useful as high-temperature circuit boards and electrical insulators.

  11. 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)

  12. Graphite-water steam-generating reactor in the USSR

    Energy Technology Data Exchange (ETDEWEB)

    Dollezhal, N A [AN SSSR, Moscow

    1981-10-01

    One of the types of power reactor used in the USSR is the graphite-water steam-generating reactor RBMK. This produces saturated steam at a pressure of 7MPa. Reactors giving 1GWe each have been installed at the Leningrad, Kursk, Chernobyl and other power stations. Further stations using reactors of this type are being built. A description is given of the fuel element design, and of the layout of the plant. The main characteristics of RBMK reactors using fuel of rated and higher enrichment are listed.

  13. Irradiation experiments of 3rd, 4th and 5th fuel assemblies by an in-pile gas loop, OGL-1

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kobayashi, Fumiaki; Hayashi, Kimio; Minato, Kazuo; Kikuchi, Teruo; Adachi, Mamoru; Iwamoto, Kazumi; Ikawa, Katsuichi; Itami, Hiroharu.

    1986-07-01

    Three irradiation experiments for 3rd, 4th and 5th fuel assemblies which had been composed of VHTR reference coated particle fuels and graphite components were carried out by an in-pile gas loop, OGL-1 during 1979 and 1982. The main purposes of these experiments were to study on bowing of the fuel rod by irradiation for the 3rd fuel assembly, to study on fuel behavior under relatively low burnup irradiation for the 4th fuel assembly, and to study on fuel behavior up to full burnup of VHTR design for the 5th fuel assembly. For understanding in-pile fuel behavior, fractional releases of fission gases from each fuel assembly were estimated by measuring the fission gas concentrations in the primary loop of OGL-1. The post-irradiation examination (PIE) was carried out extensively on the fuel block, the fuel rods and the fuel compacts in Tokai Hot Laboratory. Also, made were the measurements of metallic fission product distributions in the fuel assemblies and the fuel rods. The results in these experiments were given as follows ; bowing of the fuel rod in the 3rd fuel assembly was 0.7 mm, but integrity of the rod was kept under irradiation. Fractional release of the fission gas from the 4th fuel assembly remained in the order of 10 -7 during irradiation, suggesting that the fuel performance was excellent. The fractional release from the 5th fuel assembly, on the other hand, was in the order of 10 -5 which was the same level in the VHTR design. (author)

  14. High speed turning of compacted graphite iron using controlled modulation

    Science.gov (United States)

    Stalbaum, Tyler Paul

    Compacted graphite iron (CGI) is a material which emerged as a candidate material to replace cast iron (CI) in the automotive industry for engine block castings. Its thermal and mechanical properties allow the CGI-based engines to operate at higher cylinder pressures and temperatures than CI-based engines, allowing for lower fuel emissions and increased fuel economy. However, these same properties together with the thermomechanical wear mode in the CGI-CBN system result in poor machinability and inhibit CGI from seeing wide spread use in the automotive industry. In industry, machining of CGI is done only at low speeds, less than V = 200 m/min, to avoid encountering rapid wear of the cutting tools during cutting. Studies have suggested intermittent cutting operations such as milling suffer less severe tool wear than continuous cutting. Furthermore, evidence that a hard sulfide layer which forms over the cutting edge in machining CI at high speeds is absent during machining CGI is a major factor in the difference in machinability of these material systems. The present study addresses both of these issues by modification to the conventional machining process to allow intermittent continuous cutting. The application of controlled modulation superimposed onto the cutting process -- modulation-assisted machining (MAM) -- is shown to be quite effective in reducing the wear of cubic boron nitride (CBN) tools when machining CGI at high machining speeds (> 500 m/min). The tool life is at least 20 times greater than found in conventional machining of CGI. This significant reduction in wear is a consequence of reduction in the severity of the tool-work contact conditions with MAM. The propensity for thermochemical wear of CBN is thus reduced. It is found that higher cutting speed (> 700 m/min) leads to lower tool wear with MAM. The MAM configuration employing feed-direction modulation appears feasible for implementation at high speeds and offers a solution to this challenging

  15. Synthesis of graphitic carbon nitride by reaction of melamine and uric acid

    International Nuclear Information System (INIS)

    Dante, Roberto C.; Martin-Ramos, Pablo; Correa-Guimaraes, Adriana; Martin-Gil, Jesus

    2011-01-01

    Highlights: → Graphitic carbon nitrides by CVD of melamine and uric acid on alumina. → The building blocks of carbon nitrides are heptazine nuclei. → Composite particles with alumina core and carbon nitride coating. - Abstract: Graphitic carbon nitrides were synthesized starting from melamine and uric acid. Uric acid was chosen because it thermally decomposes, and reacts with melamine by condensation at temperatures in the range of 400-600 deg. C. The reagents were mixed with alumina and subsequently the samples were treated in an oven under nitrogen flux. Alumina favored the deposition of the graphitic carbon nitrides layers on the exposed surface. This method can be assimilated to an in situ chemical vapor deposition (CVD). Infrared (IR) spectra, as well as X-ray diffraction (XRD) patterns, are in accordance with the formation of a graphitic carbon nitride with a structure based on heptazine blocks. These carbon nitrides exhibit poor crystallinity and a nanometric texture, as shown by transmission electron microscopy (TEM) analysis. The thermal degradation of the graphitic carbon nitride occurs through cyano group formation, and involves the bridging tertiary nitrogen and the bonded carbon, which belongs to the heptazine ring, causing the ring opening and the consequent network destruction as inferred by connecting the IR and X-ray photoelectron spectroscopy (XPS) results. This seems to be an easy and promising route to synthesize graphitic carbon nitrides. Our final material is a composite made of an alumina core covered by carbon nitride layers.

  16. 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

  17. 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

  18. 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.

  19. Models development for the fuel design of a reactor GT-MHR

    International Nuclear Information System (INIS)

    Telesforo R, D.; Francois L, J. L.

    2009-10-01

    The very high temperature reactor (VHTR) it as arisen as an option for the following reactors generation, due to their characteristics as they are inherent security, modularity and relative low cost. One of the VHTR variants, for its fuel based on prismatic blocks, is the modular reactor cooled by gas GT-MHR that uses a fuel particle of multiple layers called TRISO. These particles are small containers with fuel in their interior and they function as small pressure vessels that retain the fission products. They are absorbed inconstant ing in a cylindrical graphite matrix to form the fuel named Compact; the quantity of fuel inventory is proportional to the packaging fraction on the total volume of the Compact. The reactor consists of a matrix of 12 X 12 graphite hexagonal assemblies contained in a cylinder of 3.5 radio meters and 10.0 meters high. The nucleus has ten axial regions with 36 X 3 fuel assemblies distributed in three rings. For the neutronic modeling of the fuel and the nucleus it was employee the Monte Carlo method, using the code MCNPX (Monte Carlo N-Particle version X) that is a transport code of general purposes that uses this method with a great versatility in the representation of arbitrary three-dimensional configurations and materials configuration. The heterogeneous model of the reactor GT-MHR core was obtained, adjusting the relative parameters at core prototype GT-MHR presented by General Atomics. To prove the model it was employee the fuel formed by TRISO particles, with a nucleus of 150 μm and packaging fraction of 37.55%, with a Uranium-235 mixture, as fissile nuclide, and Thorium-232. To create a simplified model of the nucleus, or homogeneous model, without modeling any particle, it was employee the reactivity-equivalent physical transformation method that captures the effects of the heterogeneity double of the fuel region in two homogeneous equivalent cells, being obtained very good results. (Author)

  20. 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)

  1. 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

  2. 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.

  3. 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

  4. 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.

  5. Some aspects of nuclear graphite production in France; Etude generale sur les graphites nucleaires produits en France

    Energy Technology Data Exchange (ETDEWEB)

    Gueron, J; Hering, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Legendre, A [Pechiney, 75 - Paris (France)

    1958-07-01

    1) Manufacturing: A summary and results on the CEA-Pechiney purification process are given. Variations in the preparation of green pastes and their effects on graphitized material are described. 2) Physical and mechanical properties: Results are given on: - Statistics of dimensional variatior products having square cross-section. - Statistical variation of thermal expansion coefficients and of electrical conductivity. - Density of normals to carbon layer planes and their connexion with thermal expansion. - Stress-strain cycles and conclusions drawn therefrom. - Mechanical resistance and gas permeability of items for supporting fuel elements. 3) Behaviour under radiation: Alteration under radiation of French graphites irradiated either in G1 pile or in experimental piles, and thermal annealing of those alterations, are given. (author)Fren. [French] 1) Fabrication: On resume le procede d'epuration CEA-PECHINEY, ainsi que diverses modalites de preparation des pates et on expose les resultats obtenus. 2) Proprietes physiques et mecaniques: On indique le resultat d'etudes sur: - la statistique des dimensions de produits a section carree. - celle des variations des coefficients de dilatation thermique et de la conductibilite electrique. - la densite des normales aux plans graphitiques et leur connexion avec la dilatation thermique. - la compression mecanique du graphite. - la solidite mecanique et la permeabilite aux gaz de pieces destinees a supporter des cartouches de combustible. 3) Tenue sous rayonnement: Modification sous rayonnement des graphites fran is irradies soit dans la pile G1, soit dans des piles experimentales, et guerison thermique de ces modifications. (auteur)

  6. Structural integrity of graphite core support structures of HTTR

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Iyoku, Tatsuo; Toyota, Junji; Sato, Sadao; Shiozawa, Shusaku

    1990-02-01

    The graphite core support structures (GCSSs) of the HTTR (High Temperature Engineering Test Reactor) are an arrangement of graphite blocks and posts that support the core and provide a lower plenum and a hot-leg path for the primary coolant. The GCSSs are designed not to be replaced by new items during plant life time (about twenty years). To maintain structural integrity of the GCSSs, conservative design has been made sufficiently on the basis of structural tests. The present study confirmed that reactor safety was still maintained even if failure and destruction of the GCSSs is supposed to occur. The GCSSs are fabricated under strict quality control and the observation and surveillance programs are planed to examine the structual integrity of the GCSSs during an operation. This paper describes the concept of design and quality control and summarizes structural tests, observation and surveillance programs. (author)

  7. The irradiation creep characteristics of graphite to high fluences

    International Nuclear Information System (INIS)

    Kennedy, C.R.; Cundy, M.; Kleist, G.

    1988-01-01

    High-temperature gas-cooled reactors (HTGR) have massive blocks of graphite with thermal and neutron-flux gradients causing high internal stresses. Thermal stresses are transient; however, stresses generated by differential growth due to neutron damage continue to increase with time. Fortunately, graphite also experiences creep under irradiation allowing relaxation of stresses to nominally safe levels. Because of complexity of irradiation creep experiments, data demonstrating this phenomenon are generally limited to fairly low fluences compared to the overall fluences expected in most reactors. Notable exceptions have been experiments at 300/degree/C and 500/degree/C run at Petten under tension and compression creep stresses to fluences greater than 4 /times/ 10 26 (E > 50 keV) neutrons/m 2 . This study complements the previous results by extending the irradiation temperature to 900/degree/C. 2 refs., 3 figs

  8. Raman characterization of bulk ferromagnetic nanostructured graphite

    International Nuclear Information System (INIS)

    Pardo, Helena; Divine Khan, Ngwashi; Faccio, Ricardo; Araújo-Moreira, F.M.; Fernández-Werner, Luciana

    2012-01-01

    Raman spectroscopy was used to characterize bulk ferromagnetic graphite samples prepared by controlled oxidation of commercial pristine graphite powder. The G:D band intensity ratio, the shape and position of the 2D band and the presence of a band around 2950 cm -1 showed a high degree of disorder in the modified graphite sample, with a significant presence of exposed edges of graphitic planes as well as a high degree of attached hydrogen atoms.

  9. Fabrication of Graphene by Cleaving Graphite Chemically

    Institute of Scientific and Technical Information of China (English)

    ZHAO Shu-hua; ZHAO Xiao-ting; FAN Hou-gang; YANG Li-li; ZHANG Yong-jun; YANG Jing-hai

    2011-01-01

    Graphite was chemically cleaved to graphene by Billups Reaction,and the morphologies and microstructures of graphene were characterized by SEM,Raman and AFM.The results show that the graphite was first functionalized by l-iodododecane,which led to the cleavage of the graphene layer in the graphite.The second decoration cleaved the graphite further and graphene was obtained.The heights of the graphene layer were larger than 1 nm due to the organic decoration.

  10. Method of Joining Graphite Fibers to a Substrate

    Science.gov (United States)

    Beringer, Durwood M. (Inventor); Caron, Mark E. (Inventor); Taddey, Edmund P. (Inventor); Gleason, Brian P. (Inventor)

    2014-01-01

    A method of assembling a metallic-graphite structure includes forming a wetted graphite subassembly by arranging one or more layers of graphite fiber material including a plurality of graphite fibers and applying a layer of metallization material to ends of the plurality of graphite fibers. At least one metallic substrate is secured to the wetted graphite subassembly via the layer of metallization material.

  11. Non-destructive evaluation on mechanical properties of nuclear graphite with porous structure

    International Nuclear Information System (INIS)

    Shibata, Taiju; Hanawa, Satoshi; Sumita, Junya; Tada, Tatsuya; Sawa, Kazuhiro; Iyoku, Tatsuo

    2005-01-01

    As a research subjects of 'Research and development for advanced high temperature gas cooled reactor fuels and graphite components,' we started the study of development of non-destructive evaluation methods for mechanical properties of graphite components. The micro-indentation and ultrasonic wave methods are focused to evaluate the degradation of graphite components in VHTR core. For the micro-indentation method, the test apparatus was designed for the indentation test on graphite specimens with some stress levels. It is expected the stress condition is evaluated by the indentation load-depth characteristics and hardness. For the ultrasonic wave method, ultrasonic wave testing machine and probes were prepared for experiments. It is expected that the stress and inner porous conditions are evaluated by the wave propagation characteristics with wave-pore interaction model. R and D plan to develop the non-destructive evaluation method for graphite is presented in this paper. (This study is the result of contract research in the fiscal year of 2004, Research and development for advanced high temperature gas cooled reactor fuels and graphite components,' which is entrusted to the Japan Atomic Energy Research Institute from the Ministry of Education, Culture, Sports, Science and Technology of Japan.) (author)

  12. ICP-MS determination of boron: method optimization during preparation of graphite reference material for boron

    International Nuclear Information System (INIS)

    Granthali, S.K.; Shailaja, P.P.; Mainsha, V.; Venkatesh, K.; Kallola, K.S.; Sanjukta, A.K.

    2017-01-01

    Graphite finds widespread use in nuclear reactors as moderator, reflector, and fuel fabricating components because of its thermal stability and integrity. The manufacturing process consists of various mixing, moulding and baking operations followed by heat-treatment between 2500 °C and 3000 °C. The high temperature treatment is required to drive the amorphous carbon-to-graphite phase transformation. Since synthetic graphite is processed at high temperature, impurity concentrations in the precursor carbon get significantly reduced due to volatilization. However boron may might partly gets converted into boron carbide at high temperatures in the carbon environment of graphite and remains stable (B_4C: boiling point 3500 °C) in the matrix. Literature survey reveals the use of various methods for determination of boron. Previously we have developed a method for determination of boron in graphite electrodes using inductively coupled plasma mass spectrometry (ICP-MS). The method involves removal of graphite matrix by ignition of the sample at 800°C in presence of saturated barium hydroxide solution to prevent the loss of boron. Here we are reporting a modification in the method by using calcium carbonate in place of barium hydroxide and using beryllium (Be) as an internal standard, which resulted in a better precession. The method was validated by spike recovery experiments as well as using another technique viz. Inductively Coupled Plasma Optical Emission Spectrometry (ICP-OES). The modified method was applied in evaluation of boron concentration in the graphite reference material prepared

  13. Photoemission study of K on graphite

    NARCIS (Netherlands)

    Bennich, P.; Puglia, C.; Brühwiler, P.A.; Nilsson, A.; Sandell, A.; Mårtensson, N.; Rudolf, P.

    1999-01-01

    The physical and electronic structure of the dispersed and (2×2) phases of K/graphite have been characterized by valence and core-level photoemission. Charge transfer from K to graphite is found to occur at all coverages, and includes transfer of charge to the second graphite layer. A rigid band

  14. Separation medium containing thermally exfoliated graphite oxide

    Science.gov (United States)

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

    2012-01-01

    A separation medium, such as a chromatography filling or packing, containing a modified graphite oxide material, which is a thermally exfoliated graphite oxide with a surface area of from about 300 m.sup.2/g to 2600 m.sup.2/g, wherein the thermally exfoliated graphite oxide has a surface that has been at least partially functionalized.

  15. NMR studies on graphite-methanol system

    International Nuclear Information System (INIS)

    El-Akkad, T.M.

    1977-01-01

    The nuclear magnetic relaxation times for protons of methanol on graphite have been studied. The perpendicular and the transversal magnetization as a function of temperature were measured. The results show that the presence of graphite slowed down the methanol movement compared with that in the pure alcohol, and that the methanol molecules are attached to the graphite surface via methyl groups. (author)

  16. The Statistical Analysis Techniques to Support the NGNP Fuel Performance Experiments

    International Nuclear Information System (INIS)

    Pham, Bihn T.; Einerson, Jeffrey J.

    2010-01-01

    This paper describes the development and application of statistical analysis techniques to support the AGR experimental program on NGNP fuel performance. The experiments conducted in the Idaho National Laboratory's Advanced Test Reactor employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule. The tests are instrumented with thermocouples embedded in graphite blocks and the target quantity (fuel/graphite temperature) is regulated by the He-Ne gas mixture that fills the gap volume. Three techniques for statistical analysis, namely control charting, correlation analysis, and regression analysis, are implemented in the SAS-based NGNP Data Management and Analysis System (NDMAS) for automated processing and qualification of the AGR measured data. The NDMAS also stores daily neutronic (power) and thermal (heat transfer) code simulation results along with the measurement data, allowing for their combined use and comparative scrutiny. The ultimate objective of this work includes (a) a multi-faceted system for data monitoring and data accuracy testing, (b) identification of possible modes of diagnostics deterioration and changes in experimental conditions, (c) qualification of data for use in code validation, and (d) identification and use of data trends to support effective control of test conditions with respect to the test target. Analysis results and examples given in the paper show the three statistical analysis techniques providing a complementary capability to warn of thermocouple failures. It also suggests that the regression analysis models relating calculated fuel temperatures and thermocouple readings can enable online regulation of experimental parameters (i.e. gas mixture content), to effectively maintain the target quantity (fuel temperature) within a given range.

  17. AMS-C14 analysis of graphite obtained with an Automated Graphitization Equipment (AGE III) from aerosol collected on quartz filters

    Energy Technology Data Exchange (ETDEWEB)

    Solís, C.; Chávez, E.; Ortiz, M.E.; Andrade, E. [Instituto de Física, Universidad Nacional Autónoma de México, 04510 México D.F. (Mexico); Ortíz, E. [Universidad Autónoma Metropolitana, Unidad Azcapotzalco, México D.F. (Mexico); Szidat, S. [Department of Chemistry and Biochemistry, University of Bern, Freiestrasse 3, CH-3012 Bern (Switzerland); Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); Wacker, L. [Laboratory of Ion Physics, ETH, Honggerberg, Zurich (Switzerland)

    2015-10-15

    AMS-{sup 14}C applications often require the analysis of small samples. Such is the case of atmospheric aerosols where frequently only a small amount of sample is available. The ion beam physics group at the ETH, Zurich, has designed an Automated Graphitization Equipment (AGE III) for routine graphite production for AMS analysis from organic samples of approximately 1 mg. In this study, we explore the potential use of the AGE III for graphitization of particulate carbon collected in quartz filters. In order to test the methodology, samples of reference materials and blanks with different sizes were prepared in the AGE III and the graphite was analyzed in a MICADAS AMS (ETH) system. The graphite samples prepared in the AGE III showed recovery yields higher than 80% and reproducible {sup 14}C values for masses ranging from 50 to 300 μg. Also, reproducible radiocarbon values were obtained for aerosol filters of small sizes that had been graphitized in the AGE III. As a study case, the tested methodology was applied to PM{sub 10} samples collected in two urban cities in Mexico in order to compare the source apportionment of biomass and fossil fuel combustion. The obtained {sup 14}C data showed that carbonaceous aerosols from Mexico City have much lower biogenic signature than the smaller city of Cuernavaca.

  18. Superconductivity in graphite intercalation compounds

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Robert P. [Cavendish Laboratory, University of Cambridge, Madingley Road, Cambridge CB3 0HE (United Kingdom); Weller, Thomas E.; Howard, Christopher A. [Department of Physics & Astronomy, University College of London, Gower Street, London WCIE 6BT (United Kingdom); Dean, Mark P.M. [Department of Condensed Matter Physics and Materials Science, Brookhaven National Laboratory, Upton, NY 11973 (United States); Rahnejat, Kaveh C. [Department of Physics & Astronomy, University College of London, Gower Street, London WCIE 6BT (United Kingdom); Saxena, Siddharth S. [Cavendish Laboratory, University of Cambridge, Madingley Road, Cambridge CB3 0HE (United Kingdom); Ellerby, Mark, E-mail: mark.ellerby@ucl.ac.uk [Department of Physics & Astronomy, University College of London, Gower Street, London WCIE 6BT (United Kingdom)

    2015-07-15

    Highlights: • Historical background of graphite intercalates. • Superconductivity in graphite intercalates and its place in the field of superconductivity. • Recent developments. • Relevant modeling of superconductivity in graphite intercalates. • Interpretations that pertain and questions that remain. - Abstract: The field of superconductivity in the class of materials known as graphite intercalation compounds has a history dating back to the 1960s (Dresselhaus and Dresselhaus, 1981; Enoki et al., 2003). This paper recontextualizes the field in light of the discovery of superconductivity in CaC{sub 6} and YbC{sub 6} in 2005. In what follows, we outline the crystal structure and electronic structure of these and related compounds. We go on to experiments addressing the superconducting energy gap, lattice dynamics, pressure dependence, and how these relate to theoretical studies. The bulk of the evidence strongly supports a BCS superconducting state. However, important questions remain regarding which electronic states and phonon modes are most important for superconductivity, and whether current theoretical techniques can fully describe the dependence of the superconducting transition temperature on pressure and chemical composition.

  19. Superconductivity in graphite intercalation compounds

    International Nuclear Information System (INIS)

    Smith, Robert P.; Weller, Thomas E.; Howard, Christopher A.; Dean, Mark P.M.; Rahnejat, Kaveh C.; Saxena, Siddharth S.; Ellerby, Mark

    2015-01-01

    Highlights: • Historical background of graphite intercalates. • Superconductivity in graphite intercalates and its place in the field of superconductivity. • Recent developments. • Relevant modeling of superconductivity in graphite intercalates. • Interpretations that pertain and questions that remain. - Abstract: The field of superconductivity in the class of materials known as graphite intercalation compounds has a history dating back to the 1960s (Dresselhaus and Dresselhaus, 1981; Enoki et al., 2003). This paper recontextualizes the field in light of the discovery of superconductivity in CaC 6 and YbC 6 in 2005. In what follows, we outline the crystal structure and electronic structure of these and related compounds. We go on to experiments addressing the superconducting energy gap, lattice dynamics, pressure dependence, and how these relate to theoretical studies. The bulk of the evidence strongly supports a BCS superconducting state. However, important questions remain regarding which electronic states and phonon modes are most important for superconductivity, and whether current theoretical techniques can fully describe the dependence of the superconducting transition temperature on pressure and chemical composition

  20. Graphite oral tattoo: case report.

    Science.gov (United States)

    Moraes, Renata Mendonça; Gouvêa Lima, Gabriela de Morais; Guilhermino, Marinaldo; Vieira, Mayana Soares; Carvalho, Yasmin Rodarte; Anbinder, Ana Lia

    2015-10-16

    Pigmented oral lesions compose a large number of pathological entities, including exogenous pigmentat oral tattoos, such as amalgam and graphite tattoos. We report a rare case of a graphite tattoo on the palate of a 62-year-old patient with a history of pencil injury, compare it with amalgam tattoos, and determine the prevalence of oral tattoos in our Oral Pathology Service. We also compare the clinical and histological findings of grafite and amalgam tattoos. Oral tattoos affect women more frequently in the region of the alveolar ridge. Graphite tattoos occur in younger patients when compared with the amalgam type. Histologically, amalgam lesions represent impregnation of the reticular fibers of vessels and nerves with silver, whereas in cases of graphite tattoos, this impregnation is not observed, but it is common to observe a granulomatous inflammatory response, less evident in cases of amalgam tattoos. Both types of lesions require no treatment, but in some cases a biopsy may be done to rule out melanocytic lesions.

  1. 'In situ' expanded graphite extinguishant

    International Nuclear Information System (INIS)

    Cao Qixin; Shou Yuemei; He Bangrong

    1987-01-01

    This report is concerning the development of the extinguishant for sodium fire and the investigation of its extinguishing property. The experiment result shows that 'in situ' expanded graphite developed by the authors is a kind of extinguishant which extinguishes sodium fire quickly and effectively and has no environment pollution during use and the amount of usage is little

  2. SNUPPS power block engineering

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, C A [Bechtel Power Corp., San Francisco, Calif. (USA)

    1975-11-01

    The Standard Power Block is based on a modular concept and consists of the following: turbine building, auxiliary building, fuel building, control building, radwaste building, diesel generators building, and outside storage tanks and transformers. Each power block unit includes a Westinghouse pressurized water reactor and has a thermal power rating of 3425 MW(t). The corresponding General Electric turbine generator net electrical output is 1188 MW(e). This standardization approach results in not only a reduction in the costs of engineering, licensing, procurement, and project planning, but should also result in additional savings by the application of experience gained in the construction of the first unit to the following units and early input of construction data to design.

  3. Graphite nanoreinforcements in polymer nanocomposites

    Science.gov (United States)

    Fukushima, Hiroyuki

    Nanocomposites composed of polymer matrices with clay reinforcements of less than 100 nm in size, are being considered for applications such as interior and exterior accessories for automobiles, structural components for portable electronic devices, and films for food packaging. While most nanocomposite research has focused on exfoliated clay platelets, the same nanoreinforcement concept can be applied to another layered material, graphite, to produce nanoplatelets and nanocomposites. Graphite is the stiffest material found in nature (Young's Modulus = 1060 GPa), having a modulus several times that of clay, but also with excellent electrical and thermal conductivity. The key to utilizing graphite as a platelet nanoreinforcement is in the ability to exfoliate this material. Also, if the appropriate surface treatment can be found for graphite, its exfoliation and dispersion in a polymer matrix will result in a composite with not only excellent mechanical properties but electrical properties as well, opening up many new structural applications as well as non-structural ones where electromagnetic shielding and high thermal conductivity are requirements. In this research, a new process to fabricate exfoliated nano-scale graphite platelets was established (Patent pending). The size of the resulted graphite platelets was less than 1 um in diameter and 10 nm in thickness, and the surface area of the material was around 100 m2/g. The reduction of size showed positive effect on mechanical properties of composites because of the increased edge area and more functional groups attached with it. Also various surface treatment techniques were applied to the graphite nanoplatelets to improve the surface condition. As a result, acrylamide grafting treatment was found to enhance the dispersion and adhesion of graphite flakes in epoxy matrices. The resulted composites showed better mechanical properties than those with commercially available carbon fibers, vapor grown carbon fibers

  4. Graphite suspension in carbon dioxide

    International Nuclear Information System (INIS)

    Roche, R.

    1965-01-01

    Since 1963 the Atomic Division of SNECMA has been conducting, under a contract with the CEA, an experimental work with a two-component fluid comprised of carbon dioxide and small graphite particles. The primary purpose was the determination of basic engineering information pertaining to the stability and the flowability of the suspension. The final form of the experimental loop consists mainly of the following items: a light-phase compressor, a heavy-phase pump, an electrical-resistance type heater section, a cooling heat exchanger, a hairpin loop, a transparent test section and a separator. During the course of the testing, it was observed that the fluid could be circulated quite easily in a broad range of variation of the suspension density and velocity - density from 30 to 170 kg/m 3 and velocity from 2 to 24 m/s. The system could be restarted and circulation maintained without any difficulty, even with the heavy-phase pump alone. The graphite did not have a tendency to pack or agglomerate during operation. No graphite deposition was observed on the wall of the tubing. A long period run (250 hours) has shown the evolution of the particle dimensions. Starting with graphite of surface area around 20 m 2 /g (graphite particles about 1 μ), the powder surface area reaches an asymptotic value of 300 m 2 /g (all the particles less than 0.3 μ). Moisture effect on flow stability, flow distribution between two parallel channels, pressure drop in straight tubes, recompression ratio in diffusers were also investigated. (author) [fr

  5. Graphite and carbonaceous materials in a molten salt nuclear reactor

    International Nuclear Information System (INIS)

    Rousseau, Ginette; Lecocq, Alfred; Hery, Michel.

    1982-09-01

    A project for a molten salt 1000 MWe reactor is studied by EDF-CEA teams. The design provides for a chromesco 3 vessel housing graphite structures in which the salt circulates. The salt (Th, U, Be and Li fluorides) is cooled by direct contact with lead. The graphites and carbonated materials, inert with respect to lead and the fuel salt, are being considered not only as moderators, but as reflectors and in the construction of the sections where the heat exchange takes place. On the basis of the problems raised in the operation of the reactor, a study programme on French experimental materials (Le Carbone Lorraine, SERS, SEP) has been defined. Hence, depending on the function or functions that the material is to ensure in the structure, the criteria of choice which follow will have to be examined: behaviour under irradiation, insertion of a fluid in the material, thermal properties required, mechanical properties required, utilization [fr

  6. CFD investigating the air ingress accident for a HTGR simulation of graphite corrosion oxidation

    International Nuclear Information System (INIS)

    Ferng, Y.M.; Chi, C.W.

    2012-01-01

    Highlights: ► A CFD model is proposed to investigate graphite oxidation corrosion in the HTR-10. ► A postulated air ingress accident is assumed in this paper. ► Air ingress flowrate is the predicted result, instead of the preset one. ► O 2 would react with graphite on pebble surface, causing the graphite corrosion. ► No fuel exposure is predicted to be occurred under the air ingress accident. - Abstract: Through a compressible multi-component CFD model, this paper investigates the characteristics of graphite oxidation corrosion in the HTR-10 core under the postulated accident of gas duct rupture. In this accident, air in the steam generator cavity would enter into the core after pressure equilibrium is achieved between the core and the cavity, which is also called as the air ingress accident. Oxygen in the air would react with graphite on pebble surface, subsequently resulting in oxidation corrosion and challenging fuel integrity. In this paper, characteristics of graphite oxidation corrosion during the air ingress accident can be reasonably captured, including distributions of graphite corrosion amount on the different cross-sections, time histories of local corrosion amount at the monitoring points and overall corrosion amount in the core, respectively. Based on the transient simulation results, the corrosion pattern and its corrosion rate would approach to the steady-state conditions as the accident continuously progresses. The total amount of graphite corrosion during a 3-day accident time is predicted to be about 31 kg with the predicted asymptotic corrosion rate. This predicted value is less than that from the previous work of Gao and Shi.

  7. 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)

  8. Identification of the key parameters defining the life of graphite core components

    International Nuclear Information System (INIS)

    Mitchell, M.N.

    2005-01-01

    The Core Structures of a Pebble Bed rector core comprise graphite reflectors constructed from blocks. These blocks are subject to high flux and temperatures as well as significant gradients in flux and temperature. This loading combined with the behaviour of graphite under irradiation gives rise to complex stress states within the reflector blocks. At some point, the stress state will reach a critical level and cracks will initiate within the blocks. The point of crack initiation is a useful point to define as the end of the part's life. The life of these graphite reflector parts in a pebble bed reactor (PBR) core determines the service life of the Core Structures. The replacement of the Core Structures' components will be a costly and time consuming. It is important that the components of the Core Structures be designed for the best life possible. As part of the conceptual design of the Pebble Bed Modular Reactor (PBMR), the assessment of the life of these components was examined. To facilitate the understanding of the parameters that influence the design life of the PBMR, a study has been completed into the effect of various design parameters on the design life of a typical side reflector block. Parameters investigated include: block geometry, material property variations, and load variations. The results of this study are to be presented. (author)

  9. Isotope heating block

    International Nuclear Information System (INIS)

    Wenk, E.

    1976-01-01

    A suggestion is made not to lead the separated nuclear 'waste' from spent nuclear fuel elements directly to end storage, but to make use of the heat produced from the remaining radiation, e.g. for seawater desalination. According to the invention, the activated fission products are to be processed, e.g. by calcination or vitrification, so that one can handle them. They should then be arranged in layers alternately with plate-shaped heat conducting pipes to form a homogeneous block; the heat absorbed by the thermal plates should be further passed on to evaporators or heat exchangers. (UWI) [de

  10. AGC-2 Graphite Preirradiation Data Package

    Energy Technology Data Exchange (ETDEWEB)

    David Swank; Joseph Lord; David Rohrbaugh; William Windes

    2012-10-01

    The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterized prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.

  11. Voronoi-Tessellated Graphite Produced by Low-Temperature Catalytic Graphitization from Renewable Resources.

    Science.gov (United States)

    Zhao, Leyi; Zhao, Xiuyun; Burke, Luke T; Bennett, J Craig; Dunlap, Richard A; Obrovac, Mark N

    2017-09-11

    A highly crystalline graphite powder was prepared from the low temperature (800-1000 °C) graphitization of renewable hard carbon precursors using a magnesium catalyst. The resulting graphite particles are composed of Voronoi-tessellated regions comprising irregular sheets; each Voronoi-tessellated region having a small "seed" particle located near their centroid on the surface. This suggests nucleated outward growth of graphitic carbon, which has not been previously observed. Each seed particle consists of a spheroidal graphite shell on the inside of which hexagonal graphite platelets are perpendicularly affixed. This results in a unique high surface area graphite with a high degree of graphitization that is made with renewable feedstocks at temperatures far below that conventionally used for artificial graphites. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  12. HTGR fuel cycle

    International Nuclear Information System (INIS)

    1987-08-01

    In the spring of 1987, the HTGR fuel cycle project has been existing for ten years, and for this reason a status seminar has been held on May 12, 1987 in the Juelich Nuclear Research Center, that gathered the participants in this project for a discussion on the state of the art in HTGR fuel element development, graphite development, and waste management. The papers present an overview of work performed so far and an outlook on future tasks and goals, and on taking stock one can say that the project has been very successful so far: The HTGR fuel element now available meets highest requirements and forms the basis of today's HTGR safety philosophy; research work on graphite behaviour in a high-temperature reactor has led to complete knowledge of the temperature or neutron-induced effects, and with the concept of direct ultimate waste disposal, the waste management problem has found a feasible solution. (orig./GL) [de

  13. Graphite structure and magnetic parameters of flake graphite cast iron

    Czech Academy of Sciences Publication Activity Database

    Vértesy, G.; Uchimoto, T.; Takagi, T.; Tomáš, Ivan; Kage, H.

    2017-01-01

    Roč. 442, Nov (2017), s. 397-402 ISSN 0304-8853 R&D Projects: GA ČR GB14-36566G Institutional support: RVO:68378271 Keywords : magnetic NDE * magnetic adaptive testing * cast iron * graphite structure * pearlite content Subject RIV: BM - Solid Matter Physics ; Magnetism OBOR OECD: Condensed matter physics (including formerly solid state physics, supercond.) Impact factor: 2.630, year: 2016

  14. Irradiation test plan of oxidation-resistant graphite in WWR-K Research Reactor

    International Nuclear Information System (INIS)

    Sumita, Junya; Shibata, Taiju; Sakaba, Nariaki; Osaki, Hirotaka; Kato, Hideki; Fujitsuka, Kunihiro; Muto, Takenori; Gizatulin, Shamil; Shaimerdenov, Asset; Dyussambayev, Daulet; Chakrov, Petr

    2014-01-01

    Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR) which is a graphite-moderated and helium gas-cooled reactor. In the case of air ingress accident in HTGR, SiO_2 protective layer is formed on the surface of SiC layer in TRISO CFP and oxidation of SiC does not proceed and fission products are retained inside the fuel particle. A new safety concept for the HTGR, called Naturally Safe HTGR, has been recently proposed. To enhance the safety of Naturally Safe HTGR ultimately, it is expected that oxidation-resistant graphite is used for graphite components to prevent the TRISO CFPs and fuel compacts from failure. SiC coating is one of candidate methods for oxidation-resistant graphite. JAEA and four graphite companies launched R&Ds to develop the oxidation-resistant graphite and the International Science and Technology Center (ISTC) partner project with JAEA and INP was launched to investigate the irradiation effects on the oxidation-resistant graphite. To determine grades of the oxidation-resistant graphite which will be adopted as irradiation test, a preliminary oxidation test was carried out. This paper described the results of the preliminary oxidation test, the plan of out-of-pile test, irradiation test and post-irradiation test (PIE) of the oxidation-resistant graphite. The results of the preliminary oxidation test showed that the integrity of the oxidation resistant graphite was confirmed and that all of grades used in the preliminary test can be adopted as the irradiation test. Target irradiation temperature was determined to be 1473 (K) and neutron fluence was determined to be from 0.54 × 10"2"5through 1.4 × 10"2"5 (/m"2, E>0.18MeV). Weight change, oxidation rate, activation energy, surface condition, etc. will be evaluated in out-of-pile test and weight change, irradiation effect on oxidation rate and activation energy, surface condition, etc. will be evaluated in PIE. (author)

  15. 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.

  16. Strategy Study on Treatment and Disposal of the Radioactive Graphite Waste of HTR

    International Nuclear Information System (INIS)

    Li Junfeng; Ma Tao; Wang Jianlong

    2014-01-01

    The possible options to change HTGR spent fuel into an acceptable form for repository disposal were discussed. The progresses of physical, chemical, and electrochemical separation of graphite from the HTGR fuel elements were reviewed. The advantages and disadvantages of each method were listed out. The total waste volume of each method was compared. The preferred option depends on the waste acceptance criteria for the repository, availability of low level waste disposal for graphite, overall economics, and overall risks. The minimum processing that yields an acceptable waste form also gives the lowest costs as well as the simplest process and the least risk. The options that could be used for treating HTGR spent fuels were listed out. The strategy for treating HTGR spent fuels and the packages needed for repository were discussed. (author)

  17. 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

    Both adsorption and absorption (sorption) of fission product (FP) gases on/into graphite are issues of interest in very high temperature reactors (VHTRs). In the original proposal, we proposed to use packed beds of graphite particles to measure sorption at a variety of temperatures and to use an electrodynamic balance (EDB) to measure sorption onto single graphite particles (a few μm in diameter) at room temperature. The use of packed beds at elevated temperature is not an issue. However, the TPOC requested revision of this initial proposal to included single particle measurements at elevated temperatures up to 1100 °C. To accommodate the desire of NEUP to extend the single particle EDB measurements to elevated temperatures it was necessary to significantly revise the plan and the budget. These revisions were approved. In the EDB method, we levitate a single graphite particle (the size, surface characteristics, morphology, purity, and composition of the particle can be varied) or agglomerate in the balance and measure the sorption of species by observing the changes in mass. This process involves the use of an electron stepping technique to measure the total charge on a particle which, in conjunction with the measured suspension voltages for the particle, allows for determinations of mass and, hence, of mass changes which then correspond to measurements of sorption. Accommodating elevated temperatures with this type of system required a significant system redesign and required additional time that ultimately was not available. These constraints also meant that the grant had to focus on fewer species as a result. Overall, the extension of the original proposed single particle work to elevated temperatures added greatly to the complexity of the proposed project and added greatly to the time that would eventually be required as well. This means that the bulk of the experimental progress was made using the packed bed sorption systems. Only being able to recruit one

  18. Ion irradiation used as surrogate of neutron irradiation in graphite: Consequences on 14C and 36Cl behavior and structural evolution

    Science.gov (United States)

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

    2018-04-01

    Graphite has been widely used as neutron moderator, reflector or fuel matrix in different types of reactors such as gas cooled nuclear reactors (UNGG, Magnox, AGR), RBMK reactors or high temperature gas cooled reactors. Their operation produces a great quantity of irradiated graphite or other carbonaceous waste (around 250,000 tons worldwide) that requires a special management strategy. In the case of disposal, which is a current management strategy, two main radionuclides, 14C and 36Cl might be dose determining at the outlet. Particular attention is paid to 14C due to its long half-life (T∼5730 years) [1] and as major contributor to the radioactive dose. 14C has two main production routes, i) transmutation of nitrogen (14N(n,p)14C) where nitrogen is mainly adsorbed at the surfaces of the irradiated graphite; ii) activation of carbon from the matrix (13C(n,γ)14C). According to leaching tests, it was shown that even if the quantity of 14C released in the solution is low (less than 1% of the initial inventory), around 30% is in the organic form that would be mobile in repository conditions [2,3]. 36Cl is mainly produced through the activation of 35Cl (35Cl(n,γ)36Cl) which is an impurity in nuclear graphite. Its activity is low but it might be highly mobile in clay host rocks. Thus, in order to make informed decisions about the best management process and to anticipate potential radionuclide dissemination during dismantling and in the repository, it is necessary to collect information on 14C and 36Cl location and speciation in graphite, after reactor closure. The goal of the present paper is therefore to use ion irradiation to simulate neutron irradiation and to evaluate the irradiation effects on the behavior of 36Cl and 14C as well as on the induced graphite structure modifications. For that, to understand and model the underlying mechanisms, we used an indirect approach based on 13C or 37Cl implantation to simulate the respective presence of 14C or 36Cl. These

  19. Optimization of temperature coefficient and breeding ratio for a graphite-moderated molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zou, C.Y.; Cai, X.Z.; Jiang, D.Z.; Yu, C.G.; Li, X.X.; Ma, Y.W.; Han, J.L. [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China); Chen, J.G., E-mail: chenjg@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China)

    2015-01-15

    Highlights: • The temperature feedback coefficient with different moderation ratios for TMSR in thermal neutron region is optimized. • The breeding ratio and doubling time of a thermal TMSR with three different reprocessing schemes are analyzed. • The smaller hexagon size and larger salt fraction with more negative feedback coefficient can better satisfy the safety demands. • A shorter reprocessing time can achieve a better breeding ratio in a thermal TMSR. • The graphite moderator lifespan is compared with other MSRs and discussed. - Abstract: Molten salt reactor (MSR) has fascinating features: inherent safety, no fuel fabrication, online fuel reprocessing, etc. However, the graphite moderated MSR may present positive feedback coefficient which has severe implications for the transient behavior during operation. In this paper, the feedback coefficient and the breeding ratio are optimized based on the fuel-to-graphite ratio variation for a thorium based MSR (TMSR). A certain thermal core with negative feedback coefficient and relative high initial breeding ratio is chosen for the reprocessing scheme analysis. The breeding performances for the TMSR under different online fuel reprocessing efficiencies and frequencies are evaluated and compared with other MSR concepts. The results indicate that the thermal TMSR can get a breeding ratio greater than 1.0 with appropriate reprocessing scheme. The low fissile inventory in thermal TMSR leads to a short doubling time and low transuranic (TRU) inventory. The lifetime of graphite used for the TMSR is also discussed.

  20. AGC-3 Graphite Preirradiation Data Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    William Windes; David Swank; David Rohrbaugh; Joseph Lord

    2013-09-01

    This report describes the specimen loading order and documents all pre-irradiation examination material property measurement data for the graphite specimens contained within the third Advanced Graphite Capsule (AGC-3) irradiation capsule. The AGC-3 capsule is third in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens allowing quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades to be generated which will ascertain the in-service behavior of the graphite for pebble bed and prismatic Very High Temperature Reactor (VHTR) designs. The general design of AGC-3 test capsule is similar to the AGC-2 test capsule, material property tests were conducted on graphite specimens prior to loading into the AGC-3 irradiation assembly. However the 6 major nuclear graphite grades in AGC-2 were modified; two previous graphite grades (IG-430 and H-451) were eliminated and one was added (Mersen’s 2114 was added). Specimen testing from three graphite grades (PCEA, 2114, and NBG-17) was conducted at Idaho National Laboratory (INL) and specimen testing for two grades (IG-110 and NBG-18) were conducted at Oak Ridge National Laboratory (ORNL) from May 2011 to July 2013. This report also details the specimen loading methodology for the graphite specimens inside the AGC-3 irradiation capsule. The AGC-3 capsule design requires "matched pair" creep specimens that have similar dose levels above and below the neutron flux profile mid-plane to provide similar specimens with and without an applied load. This document utilized the neutron flux profile calculated for the AGC-3 capsule design, the capsule dimensions, and the size (length) of the selected graphite and silicon carbide samples to create a stacking order that can produce "matched pairs" of graphite samples above and below the AGC-3 capsule elevation mid-point to

  1. Analysis of triso packing fraction and fissile material to DB-MHR using LWR reprocessed fuel

    International Nuclear Information System (INIS)

    Silva, Clarysson A.M. da; Pereira, Claubia; Costa, Antonella L.; Veloso, Maria Auxiliadora F.; Gual, Maritza R.

    2013-01-01

    Gas-cooled and graphite-moderated reactor is being considered the next generation of nuclear power plants because of its characteristic to operate with reprocessed fuel. The typical fuel element consists of a hexagonal block with coolant and fuel channels. The fuel pin is manufactured into compacted ceramic-coated particles (TRISO) which are used to achieve both a high burnup and a high degree of passive safety. This work uses the MCNPX 2.6.0 to simulate the active core of Deep Burn Modular Helium Reactor (DB-MHR) employing PWR (Pressurized Water Reactor) reprocessed fuel. However, before a complete study of DB-MHR fuel cycle and recharge, it is necessary to evaluate the neutronic parameters to some values of TRISO Packing Fractions (PF) and Fissile Material (FM). Each PF and FM combination would generate the best behaviour of neutronic parameters. Therefore, this study configures several PF and FM combinations considering the heterogeneity of TRISO layers and lattice. The results present the best combination of PF and FM values according with the more appropriated behaviour of the neutronic parameters during the burnup. In this way, the optimized combination can be used to future works of MHR fuel cycle and recharge. (author)

  2. Leaching of 14C and 36Cl from irradiated French graphite

    International Nuclear Information System (INIS)

    Gray, W.J.; Morgan, W.C.

    1989-08-01

    The leach rates of 14 C and 36 Cl were measured on solid cylindrical samples prepared from irradiated graphite blocks supplied by the French Commissariat a l'Energie Atomique (CEA). Static leach tests were conducted in deionized water at 20 degree C for 13 weeks. The graphite samples were completely submerged in the water, and the entire volume of water was changed and analyzed at weekly intervals for the first three weeks and biweekly thereafter. Large differences in the leach rates of both 14 C and 36 Cl were observed between samples machined from the different blocks. In general, the leach rates were much higher than those measured in an earlier study with graphite obtained from a block removed from one of the Hanford reactors. The data from this study are compared with those from the previous study using the Hanford-reactor graphite. Implications of the data from both studies regarding possible rate-limiting mechanisms are discussed. 4 refs., 8 figs., 3 tabs

  3. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part II: Prismatic Reactor Cross Section Generation

    Energy Technology Data Exchange (ETDEWEB)

    Vincent Descotes

    2011-03-01

    The deep-burn prismatic high temperature reactor is made up of an annular core loaded with transuranic isotopes and surrounded in the center and in the periphery by reflector blocks in graphite. This disposition creates challenges for the neutronics compared to usual light water reactor calculation schemes. The longer mean free path of neutrons in graphite affects the neutron spectrum deep inside the blocks located next to the reflector. The neutron thermalisation in the graphite leads to two characteristic fission peaks at the inner and outer interfaces as a result of the increased thermal flux seen in those assemblies. Spectral changes are seen at least on half of the fuel blocks adjacent to the reflector. This spectral effect of the reflector may prevent us from successfully using the two step scheme -lattice then core calculation- typically used for light water reactors. We have been studying the core without control mechanisms to provide input for the development of a complete calculation scheme. To correct the spectrum at the lattice level, we have tried to generate cross-sections from supercell calculations at the lattice level, thus taking into account part of the graphite surrounding the blocks of interest for generating the homogenised cross-sections for the full-core calculation. This one has been done with 2 to 295 groups to assess if increasing the number of groups leads to more accurate results. A comparison with a classical single block model has been done. Both paths were compared to a reference calculation done with MCNP. It is concluded that the agreement with MCNP is better with supercells, but that the single block model remains quite close if enough groups are kept for the core calculation. 26 groups seems to be a good compromise between time and accu- racy. However, some trials with depletion have shown huge variations of the isotopic composition across a block next to the reflector. It may imply that at least an in- core depletion for the

  4. Environmentally benign graphite intercalation compound composition for exfoliated graphite, flexible graphite, and nano-scaled graphene platelets

    Science.gov (United States)

    Zhamu, Aruna; Jang, Bor Z.

    2014-06-17

    A carboxylic-intercalated graphite compound composition for the production of exfoliated graphite, flexible graphite, or nano-scaled graphene platelets. The composition comprises a layered graphite with interlayer spaces or interstices and a carboxylic acid residing in at least one of the interstices, wherein the composition is prepared by a chemical oxidation reaction which uses a combination of a carboxylic acid and hydrogen peroxide as an intercalate source. Alternatively, the composition may be prepared by an electrochemical reaction, which uses a carboxylic acid as both an electrolyte and an intercalate source. Exfoliation of the invented composition does not release undesirable chemical contaminants into air or drainage.

  5. Electrochemical Ultracapacitors Using Graphitic Nanostacks

    Science.gov (United States)

    Marotta, Christopher

    2012-01-01

    Electrochemical ultracapacitors (ECs) have been developed using graphitic nanostacks as the electrode material. The advantages of this technology will be the reduction of device size due to superior power densities and relative powers compared to traditional activated carbon electrodes. External testing showed that these materials display reduced discharge response times compared to state-of-the-art materials. Such applications are advantageous for pulsed power applications such as burst communications (satellites, cell phones), electromechanical actuators, and battery load leveling in electric vehicles. These carbon nanostructures are highly conductive and offer an ordered mesopore network. These attributes will provide more complete electrolyte wetting, and faster release of stored charge compared to activated carbon. Electrochemical capacitor (EC) electrode materials were developed using commercially available nanomaterials and modifying them to exploit their energy storage properties. These materials would be an improvement over current ECs that employ activated carbon as the electrode material. Commercially available graphite nanofibers (GNFs) are used as precursor materials for the synthesis of graphitic nanostacks (GNSs). These materials offer much greater surface area than graphite flakes. Additionally, these materials offer a superior electrical conductivity and a greater average pore size compared to activated carbon electrodes. The state of the art in EC development uses activated carbon (AC) as the electrode material. AC has a high surface area, but its small average pore size inhibits electrolyte ingress/egress. Additionally, AC has a higher resistivity, which generates parasitic heating in high-power applications. This work focuses on fabricating EC from carbon that has a very different structure by increasing the surface area of the GNF by intercalation or exfoliation of the graphitic basal planes. Additionally, various functionalities to the GNS

  6. Graphite materials testing in the ATR for lifetime management of Magnox reactors

    International Nuclear Information System (INIS)

    Grover, S.B.; Metcalfe, M.P.

    2002-01-01

    A major feature of the Magnox gas cooled reactor design is the graphite core, which acts as the moderator but also provides the physical structure for fuel, control rods, instrumentation and coolant gas channels. The lifetime of a graphite core is dependent upon two principal aging processes: irradiation damage and radiolytic oxidation. Irradiation damage from fast neutrons creates lattice defects leading to changes in physical and mechanical properties and the accumulation of stresses. Radiolytic oxidation is caused by the reaction of oxidizing species from the carbon dioxide coolant gas with the graphite, these species being produced by gamma radiation. Radiolytic oxidation reduces the density and hence the moderating capability of the graphite, but also reduces strength affecting the integrity of core components. In order to manage continued operation over the planned lifetimes of their power stations, BNFL needed to extend their database of the effects of these two phenomena on their graphite cores through an irradiation experiment. This paper will discuss the background, purpose, and the processes taken and planned (i.e. post irradiation examination) to ensure meaningful data on the graphite core material is obtained from the irradiation experiment. (author)

  7. Graphite Materials Testing in the ATR for Lifetime Management of Magnox Reactors

    International Nuclear Information System (INIS)

    Grover, S.B.; Metcalfe, M.P.

    2002-01-01

    A major feature of the Magnox gas cooled reactor design is the graphite core, which acts as the moderator but also provides the physical structure for fuel, control rods, instrumentation and coolant gas channels. The lifetime of a graphite core is dependent upon two principal aging processes: irradiation damage and radiolytic oxidation. Irradiation damage from fast neutrons creates lattice defects leading to changes in physical and mechanical properties and the accumulation of stresses. Radiolytic oxidation is caused by the reaction of oxidizing species from the carbon dioxide coolant gas with the graphite, these species being produced by gamma radiation. Radiolytic oxidation reduces the density and hence the moderating capability of the graphite, but also reduces strength affecting the integrity of core components. In order to manage continued operation over the planned lifetimes of their power stations, BNFL needed to extend their database of the effects of these two phenomena on the ir graphite cores through an irradiation experiment. This paper will discuss the background, purpose, and the processes taken and planned (i.e. post irradiation examination) to ensure meaningful data on the graphite core material is obtained from the irradiation experiment

  8. 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)

  9. Pyrolytic graphite gauge for measuring heat flux

    Science.gov (United States)

    Bunker, Robert C. (Inventor); Ewing, Mark E. (Inventor); Shipley, John L. (Inventor)

    2002-01-01

    A gauge for measuring heat flux, especially heat flux encountered in a high temperature environment, is provided. The gauge includes at least one thermocouple and an anisotropic pyrolytic graphite body that covers at least part of, and optionally encases the thermocouple. Heat flux is incident on the anisotropic pyrolytic graphite body by arranging the gauge so that the gauge surface on which convective and radiative fluxes are incident is perpendicular to the basal planes of the pyrolytic graphite. The conductivity of the pyrolytic graphite permits energy, transferred into the pyrolytic graphite body in the form of heat flux on the incident (or facing) surface, to be quickly distributed through the entire pyrolytic graphite body, resulting in small substantially instantaneous temperature gradients. Temperature changes to the body can thereby be measured by the thermocouple, and reduced to quantify the heat flux incident to the body.

  10. Attenuation of thermal neutron through graphite

    International Nuclear Information System (INIS)

    Adib, M.; Ismaail, H.; Fathaallah, M.; Abbas, Y.; Habib, N.; Wahba, M.

    2004-01-01

    Calculation of the nuclear capture, thermal diffuse and Bragg scattering cross-sections as a function of graphite temperature and crystalline from for neutron energies from 1 me V< E<10 eV were carried out. Computer programs have been developed which allow calculation for the graphite hexagonal closed-pack structure in its polycrystalline form and pyrolytic one. I The calculated total cross-section for polycrystalline graphite were compared with the experimental values. An overall agreement is indicated between the calculated values and experimental ones. Agreement was also obtained for neutron cross-section measured for oriented pyrolytic graphite at room and liquid nitrogen temperatures. A feasibility study for use of graphite in powdered form as a cold neutron filter is details. The calculated attenuation of thermal neutrons through large mosaic pyrolytic graphite show that such crystals can be used effectively as second order filter of thermal neutron beams and that cooling improve their effectiveness

  11. Fuel element for high-temperature nuclear power reactors

    International Nuclear Information System (INIS)

    Schloesser, J.

    1974-01-01

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

  12. Tungsten Deposition on Graphite using Plasma Enhanced Chemical Vapour Deposition

    International Nuclear Information System (INIS)

    Sharma, Uttam; Chauhan, Sachin S; Sharma, Jayshree; Sanyasi, A K; Ghosh, J; Choudhary, K K; Ghosh, S K

    2016-01-01

    The tokamak concept is the frontrunner for achieving controlled thermonuclear reaction on earth, an environment friendly way to solve future energy crisis. Although much progress has been made in controlling the heated fusion plasmas (temperature ∼ 150 million degrees) in tokamaks, technological issues related to plasma wall interaction topic still need focused attention. In future, reactor grade tokamak operational scenarios, the reactor wall and target plates are expected to experience a heat load of 10 MW/m 2 and even more during the unfortunate events of ELM's and disruptions. Tungsten remains a suitable choice for the wall and target plates. It can withstand high temperatures, its ductile to brittle temperature is fairly low and it has low sputtering yield and low fuel retention capabilities. However, it is difficult to machine tungsten and hence usages of tungsten coated surfaces are mostly desirable. To produce tungsten coated graphite tiles for the above-mentioned purpose, a coating reactor has been designed, developed and made operational at the SVITS, Indore. Tungsten coating on graphite has been attempted and successfully carried out by using radio frequency induced plasma enhanced chemical vapour deposition (rf -PECVD) for the first time in India. Tungsten hexa-fluoride has been used as a pre-cursor gas. Energy Dispersive X-ray spectroscopy (EDS) clearly showed the presence of tungsten coating on the graphite samples. This paper presents the details of successful operation and achievement of tungsten coating in the reactor at SVITS. (paper)

  13. Dynamics of graphite flake on a liquid

    Science.gov (United States)

    Miura, K.; Tsuda, D.; Kaneta, Y.; Harada, R.; Ishikawa, M.; Sasaki, N.

    2006-11-01

    One-directional motion, where graphite flakes are driven by a nanotip on an octamethylcyclotetrasiloxane (OMCTS) liquid surface, is presented. A transition from quasiperiodic to chaotic motions occurs in the dynamics of a graphite flake when its velocity is increased. The dynamics of graphite flakes pulled by the nanotip on an OMCTS liquid surface can be treated as that of a nanobody on a liquid.

  14. Sealing nuclear graphite with pyrolytic carbon

    International Nuclear Information System (INIS)

    Feng, Shanglei; Xu, Li; Li, Li; Bai, Shuo; Yang, Xinmei; Zhou, Xingtai

    2013-01-01

    Pyrolytic carbon (PyC) coatings were deposited on IG-110 nuclear graphite by thermal decomposition of methane at ∼1830 °C. The PyC coatings are anisotropic and airtight enough to protect IG-110 nuclear graphite against the permeation of molten fluoride salts and the diffusion of gases. The investigations indicate that the sealing nuclear graphite with PyC coating is a promising method for its application in Molten Salt Reactor (MSR)

  15. 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)

  16. Nanostructured carbon films with oriented graphitic planes

    International Nuclear Information System (INIS)

    Teo, E. H. T.; Kalish, R.; Kulik, J.; Kauffmann, Y.; Lifshitz, Y.

    2011-01-01

    Nanostructured carbon films with oriented graphitic planes can be deposited by applying energetic carbon bombardment. The present work shows the possibility of structuring graphitic planes perpendicular to the substrate in following two distinct ways: (i) applying sufficiently large carbon energies for deposition at room temperature (E>10 keV), (ii) utilizing much lower energies for deposition at elevated substrate temperatures (T>200 deg. C). High resolution transmission electron microscopy is used to probe the graphitic planes. The alignment achieved at elevated temperatures does not depend on the deposition angle. The data provides insight into the mechanisms leading to the growth of oriented graphitic planes under different conditions.

  17. Production of nuclear graphite in France

    International Nuclear Information System (INIS)

    Legendre, P.; Mondet, L.; Arragon, Ph.; Cornuault, P.; Gueron, J.; Hering, H.

    1955-01-01

    The graphite intended for the construction of the reactors is obtained by the usual process: confection of a cake from coke of oil and tar, cooked (in a electric oven) then the product of cook is graphitized, also by electric heating. The use of the air transportation and the control of conditions cooking and graphitization have permitted to increase the nuclear graphite production as well as to better control their physical and mechanical properties and to reduce to the minimum the unwanted stains. (M.B.) [fr

  18. AC induction field heating of graphite foam

    Science.gov (United States)

    Klett, James W.; Rios, Orlando; Kisner, Roger

    2017-08-22

    A magneto-energy apparatus includes an electromagnetic field source for generating a time-varying electromagnetic field. A graphite foam conductor is disposed within the electromagnetic field. The graphite foam when exposed to the time-varying electromagnetic field conducts an induced electric current, the electric current heating the graphite foam. An energy conversion device utilizes heat energy from the heated graphite foam to perform a heat energy consuming function. A device for heating a fluid and a method of converting energy are also disclosed.

  19. Structural analysis of polycrystalline (graphitized) materials

    International Nuclear Information System (INIS)

    Efremenko, M.M.; Kravchik, A.E.; Osmakov, A.S.

    1993-01-01

    Specific features of the structure of polycrystal carbon materials (CM), characterized by high enough degree of structural perfection and different genesis are analyzed. From the viewpoint of fine and supercrystallite structure analysis of the most characteristic groups of graphitized CM: artificial graphites, and natural graphites, as well, has been carried out. It is ascertained that in paracrystal CM a monolayer of hexagonally-bound carbon atoms is the basic element of the structure, and in graphitized CM - a microlayer. The importance of the evaluation of the degree of three-dimensional ordering of the microlayer is shown

  20. Low temperature vapor phase digestion of graphite

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, Robert A.

    2017-04-18

    A method for digestion and gasification of graphite for removal from an underlying surface is described. The method can be utilized to remove graphite remnants of a formation process from the formed metal piece in a cleaning process. The method can be particularly beneficial in cleaning castings formed with graphite molding materials. The method can utilize vaporous nitric acid (HNO.sub.3) or vaporous HNO.sub.3 with air/oxygen to digest the graphite at conditions that can avoid damage to the underlying surface.

  1. The Fracture Toughness of Nuclear Graphites Grades

    Energy Technology Data Exchange (ETDEWEB)

    Burchell, Timothy D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Erdman, III, Donald L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lowden, Rick R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunter, James A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hannel, Cara C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-04-01

    New measurements of graphite mode I critical stress intensity factor, KIc (commonly referred to as the fracture toughness) and the mode II critical shear stress intensity, KIIc, are reported and compared with prior data for KIc and KIIc. The new data are for graphite grades PCEA, IG-110 and 2114. Variations of KIc and acoustic emission (AE) data with graphite texture are reported and discussed. The Codes and Standards applications of fracture toughness, KIc, data are also discussed. A specified minimum value for nuclear graphite KIc is recommended.

  2. Temperature Analysis and Failure Probability of the Fuel Element in HTR-PM

    International Nuclear Information System (INIS)

    Yang Lin; Liu Bing; Tang Chunhe

    2014-01-01

    Spherical fuel element is applied in the 200-MW High Temperature Reactor-Pebble-bed Modular (HTR-PM). Each spherical fuel element contains approximately 12,000 coated fuel particles in the inner graphite matrix with a diameter of 50mm to form the fuel zone, while the outer shell with a thickness of 5mm is a fuel-free zone made up of the same graphite material. Under high burnup irradiation, the temperature of fuel element rises and the stress will result in the damage of fuel element. The purpose of this study is to analyze the temperature of fuel element and to discuss the stress and failure probability. (author)

  3. Electrolysis of acidic sodium chloride solution with a graphite anode. I. Graphite electrode

    NARCIS (Netherlands)

    Janssen, L.J.J.; Hoogland, J.G.

    1969-01-01

    A graphite anode evolving Cl from a chloride soln. is slowly oxidized to CO and CO2. This oxidn. causes a change in the characteristics of the electrode in aging, comprising a change of the nature of the graphite surface and an increase of the surface area. It appears that a new graphite electrode

  4. An efficient, block-by-block algorithm for inverting a block tridiagonal, nearly block Toeplitz matrix

    International Nuclear Information System (INIS)

    Reuter, Matthew G; Hill, Judith C

    2012-01-01

    We present an algorithm for computing any block of the inverse of a block tridiagonal, nearly block Toeplitz matrix (defined as a block tridiagonal matrix with a small number of deviations from the purely block Toeplitz structure). By exploiting both the block tridiagonal and the nearly block Toeplitz structures, this method scales independently of the total number of blocks in the matrix and linearly with the number of deviations. Numerical studies demonstrate this scaling and the advantages of our method over alternatives.

  5. Emplacement of small and large buffer blocks

    International Nuclear Information System (INIS)

    Saari, H.; Nikula, M.; Suikki, M.

    2010-05-01

    The report describes emplacement of a buffer structure encircling a spent fuel canister to be deposited in a vertical hole. The report deals with installability of various size blocks and with an emplacement gear, as well as evaluates the achieved quality of emplacement and the time needed for installing the buffer. Two block assembly of unequal size were chosen for examination. A first option involved small blocks, the use of which resulted in a buffer structure consisting of small sector blocks 200 mm in height. A second option involved large blocks, resulting in a buffer structure which consists of eight blocks. In these tests, the material chosen for both block options was concrete instead of bentonite. The emplacement test was a three-phase process. A first phase included stacking a two meter high buffer structure with small blocks for ensuring the operation of test equipment and blocks. A second phase included installing buffer structures with both block options to a height matching that of a canister-encircling cylindrical component. A third phase included testing also the installability of blocks to be placed above the canister by using small blocks. In emplacement tests, special attention was paid to the installability of blocks as well as to the time required for emplacement. Lifters for both blocks worked well. Due to the mass to be lifted, the lifter for large blocks had a more heavy-duty frame structure (and other lifting gear). The employed lifters were suspended in the tests on a single steel wire rope. Stacking was managed with both block sizes at adequate precision and stacked-up towers were steady. The stacking of large blocks was considerably faster. Therefore it is probably that the overall handling of the large blocks will be more convenient at a final disposal site. From the standpoint of reliability in lifting, the small blocks were safer to install above the canister. In large blocks, there are strict shape-related requirements which are

  6. Nitrogen-doped graphene by ball-milling graphite with melamine for energy conversion and storage

    International Nuclear Information System (INIS)

    Xue, Yuhua; Chen, Hao; Qu, Jia; Dai, Liming

    2015-01-01

    N-doped graphene was prepared by ball milling of graphite with melamine. It was found that ball-milling reduced the size of graphite particles from 30 to 1 μm and facilitated the exfoliation of the resultant small particles into few-layer N-doped graphene nanosheets under ultrasonication. The as-prepared N-doped graphene nanoplatelets (NGnPs) exhibited a nitrogen content as high as 11.4 at.%, making them attractive as efficient electrode materials in supercapacitors for energy storage and as highly-active metal-free catalysts for oxygen reduction in fuel cells for energy conversion. (paper)

  7. Advances in HTGR spent fuel treatment technology

    International Nuclear Information System (INIS)

    Holder, N.D.; Lessig, W.S.

    1984-08-01

    GA Technologies, Inc. has been investigating the burning of spent reactor graphite under Department of Energy sponsorship since 1969. Several deep fluidized bed burners have been used at the GA pilot plant to develop graphite burning techniques for both spent fuel recovery and volume reduction for waste disposal. Since 1982 this technology has been extended to include more efficient circulating bed burners. This paper includes updates on high-temperature gas-cooled reactor fuel cycle options and current results of spent fuel treatment testing for fluidized and advanced circulating bed burners

  8. Analysis of Block OMP using Block RIP

    OpenAIRE

    Wang, Jun; Li, Gang; Zhang, Hao; Wang, Xiqin

    2011-01-01

    Orthogonal matching pursuit (OMP) is a canonical greedy algorithm for sparse signal reconstruction. When the signal of interest is block sparse, i.e., it has nonzero coefficients occurring in clusters, the block version of OMP algorithm (i.e., Block OMP) outperforms the conventional OMP. In this paper, we demonstrate that a new notion of block restricted isometry property (Block RIP), which is less stringent than standard restricted isometry property (RIP), can be used for a very straightforw...

  9. Mixed graphite cast iron for automotive exhaust component applications

    OpenAIRE

    De-lin Li

    2017-01-01

    Both spheroidal graphite iron and compacted graphite iron are used in the automotive industry. A recently proposed mixed graphite iron exhibits a microstructure between the conventional spheroidal graphite iron and compacted graphite iron. Evaluation results clearly indicate the suitability and benefits of mixed graphite iron for exhaust component applications with respect to casting, machining, mechanical, thermophysical, oxidation, and thermal fatigue properties. A new ASTM standard speci...

  10. Nuclear graphite waste management. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2001-05-01

    The purpose of the seminar was to bring together the specialists dealing with various aspects of radioactive graphite waste management to exchange and review information on the decommissioning, characterisation, processing and disposal of irradiated graphite from reactor cores and other graphite waste associated with reactor operation. The seminar covered radioactive graphite characterisation, the effect of irradiation on graphite components, Wigner energy, radioactive graphite waste treatment, conditioning, interim storage and long term disposal options. Individual papers presented at the seminar were indexed separately

  11. Clean energy from a carbon fuel cell

    Science.gov (United States)

    Kacprzak, Andrzej; Kobyłecki, Rafał; Bis, Zbigniew

    2011-12-01

    The direct carbon fuel cell technology provides excellent conditions for conversion of chemical energy of carbon-containing solid fuels directly into electricity. The technology is very promising since it is relatively simple compared to other fuel cell technologies and accepts all carbon-reach substances as possible fuels. Furthermore, it makes possible to use atmospheric oxygen as the oxidizer. In this paper the results of authors' recent investigations focused on analysis of the performance of a direct carbon fuel cell supplied with graphite, granulated carbonized biomass (biocarbon), and granulated hard coal are presented. The comparison of the voltage-current characteristics indicated that the results obtained for the case when the cell was operated with carbonized biomass and hard coal were much more promising than those obtained for graphite. The effects of fuel type and the surface area of the cathode on operation performance of the fuel cell were also discussed.

  12. 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.

  13. Mixed graphite cast iron for automotive exhaust component applications

    Directory of Open Access Journals (Sweden)

    De-lin Li

    2017-11-01

    Full Text Available Both spheroidal graphite iron and compacted graphite iron are used in the automotive industry. A recently proposed mixed graphite iron exhibits a microstructure between the conventional spheroidal graphite iron and compacted graphite iron. Evaluation results clearly indicate the suitability and benefits of mixed graphite iron for exhaust component applications with respect to casting, machining, mechanical, thermophysical, oxidation, and thermal fatigue properties. A new ASTM standard specification (A1095 has been created for compacted, mixed, and spheroidal graphite silicon-molybdenum iron castings. This paper attempts to outline the latest progress in mixed graphite iron published.

  14. Methodology of characterization of radioactive graphite

    International Nuclear Information System (INIS)

    Pina, G.; Rodriguez, M.; Lara, E.; Magro, E.; Gascon, J. L.; Leganes, J. L.

    2014-01-01

    Since the dismantling of Vandellos I, ENRESA has promoted the precise knowledge of the inventory of irradiated graphite (graphite-i) through establishing methodologies for radiological characterization of the vector of radionuclides of interest and their correlations as the primary means of characterization strategy to establish the safer management of this material in its life cycle. (Author)

  15. Significance of primary irradiation creep in graphite

    CSIR Research Space (South Africa)

    Erasmus, C

    2013-05-01

    Full Text Available Traditionally primary irradiation creep is introduced into graphite analysis by applying the appropriate amount of creep strain to the model at the initial time-step. This is valid for graphite components that are subjected to high fast neutron flux...

  16. 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)

  17. Metal/graphite - composites in fusion engineering

    International Nuclear Information System (INIS)

    Staffler, R.; Kneringer, G.; Kny, E.; Reheis, N.

    1989-01-01

    Metal/graphite composites have been well known in medical industry for many years. X-ray tubes used in modern radiography, particularly in computerized tomography are equipped with rotating targets able to absorb a maximum of heat in a given time. Modern rotating targets consist of a refractory metal/graphite composite. Today the use of graphite as a plasma facing material is one predominant concept in fusion engineering. Depending on the thermal load, the graphite components have to be directly cooled (i.e. divertor plates) or inertially cooled (i.e. firstwall tiles). In case of direct cooling a metallurgical joining such as high temperature brazing between graphite and a metallic cooling structure shows the most promising results /1/. Inertially cooled graphite tiles have to be joined to a metallic backing plate in order to get a stable attachment to the supporting structure. The main requirements on the metallic partner of a metal/graphite composite used in the first wall area are: high melting point, high thermal strength, high thermal conductivity, low vapor pressure and a thermal expansion matching that of graphite. These properties are typical for the refractory metals such as molybdenum, tungsten and their alloys. 4 refs., 13 figs., 1 tab

  18. Metal/graphite - composites in fusion engineering

    International Nuclear Information System (INIS)

    Staffler, R.; Kneringer, G.; Kny, E.; Reheis, N.

    1995-01-01

    Metal/graphite composites have been well known in medical industry for many years. X-ray tubes used in modern radiography, particulary in computerized tomography are equipped with rotating targets able to absorb a maximum of heat in a given time. Modern rotating targets consist of a refractory metal/graphite composite. Today the use of graphite as a plasma facing material is one predominant concept in fusion engineering. Depending on the thermal load, the graphite components have to be directly cooled (i.e. divertor plates) or inertially cooled (i.e. firstwall tiles). In case of direct cooling a metallurgical joining such as high temperature brazing between graphite and a metalic cooling structure shows the most promising results /1/. Inertially cooled graphite tiles have to be joined to a metallic backing plate in order to get a stable attachment to the supporting structure. The main requirements on the metallic partner of a metal/graphite composite and in the first wall area are: high melting point, high thermal strength, high thermal conductivity, low vapour pressure and a thermal expansion matching that of graphite. These properties are typical for the refractory metals such as molybdenum, tungsten and their alloys. (author)

  19. Tire containing thermally exfoliated graphite oxide

    Science.gov (United States)

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

    2011-01-01

    A tire, tire lining or inner tube, containing a polymer composite, made of at least one rubber and/or at least one elastomer and a modified graphite oxide material, which is a thermally exfoliated graphite oxide with a surface area of from about 300 sq m/g to 2600 sq m/g.

  20. Effect of graphite target power density on tribological properties of graphite-like carbon films

    Science.gov (United States)

    Dong, Dan; Jiang, Bailing; Li, Hongtao; Du, Yuzhou; Yang, Chao

    2018-05-01

    In order to improve the tribological performance, a series of graphite-like carbon (GLC) films with different graphite target power densities were prepared by magnetron sputtering. The valence bond and microstructure of films were characterized by AFM, TEM, XPS and Raman spectra. The variation of mechanical and tribological properties with graphite target power density was analyzed. The results showed that with the increase of graphite target power density, the deposition rate and the ratio of sp2 bond increased obviously. The hardness firstly increased and then decreased with the increase of graphite target power density, whilst the friction coefficient and the specific wear rate increased slightly after a decrease with the increasing graphite target power density. The friction coefficient and the specific wear rate were the lowest when the graphite target power density was 23.3 W/cm2.

  1. 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.

  2. CANDU fuel performance and development

    International Nuclear Information System (INIS)

    Hardy, D.G.; Wood, J.C.; Bain, A.S.

    1978-12-01

    The fuel defect rate in CANDU (Canada Deuterium Uranium) reactors continues to be very low, 0.06% since 1972. The power ramp defects, which constituted the majority of the early defects, have been virtually eliminated by changed fuelling schemes and through the introduction of graphite CANLUB coatings on the inside of the sheath. Laboratory and loop irradiations have demonstrated that the graphite CANLUB layers increase the tolerance to power ramps, but to obtain the maximum benefit, coating parameters such as thickness, adhesion and wear resistance must be optimized. Siloxane CANLUB coated fuel offers greater tolerance to power ramps than most graphite coatings; quality control appears simpler and no instance of localized sheath hydriding has been seen with cured and irradiated coatings. Limited testing has shown that fuel with graphite discs between fuel pellets also has high tolerance to power ramps, but it is more costly and has lower burnup. The number of defects due to faulty components has been extremely small (0.00014%), but improved quality control and welding procedures can lower this number even further. Defects from causes external to the bundle have also been very few. (author)

  3. The role of graphite foam pore structure on saturated pool boiling enhancement

    International Nuclear Information System (INIS)

    Pranoto, I.; Leong, K.C.; Jin, L.W.

    2012-01-01

    This paper presents an experimental study of the pool boiling phenomena and performance of porous graphite foam evaporators of different structures and thermophysical properties. Two dielectric liquids viz. FC-72 and HFE-7000 were used as working fluids. Block and fin evaporators of different fin-to-block-surface-area ratios (AR) were designed to study the role of the internal pore structure of graphite foams in a compact air-cooled thermosyphon under saturated pool boiling condition for high heat flux electronics cooling applications. The wall temperatures were measured and the boiling heat transfer coefficients were calculated to analyze the boiling performance. It was found that both fin structures with AR = 3.70 and 2.73 result in reduced boiling heat transfer performances and higher wall temperatures. The experimental results show that the boiling heat transfer coefficients of the block structures are about 1.2–1.6 times higher than those of the fin structures. The total internal surface area to volume ratio (β) and the total exposed areas (A T ) of the graphite foams were calculated in this study. The results show that the values of β and A T of the block structures are much higher than the fin structures for both tested “Pocofoam” 61% porosity and “Kfoam” 78% porosity evaporators which resulted in higher boiling heat transfer coefficient and lower wall temperature of the block structures. A visualization study shows that more bubbles were generated from the block structures compared to the fin structures due to the larger number of nucleation sites from the block structures. It was also found that use of FC-72 resulted in better boiling heat transfer performance compared to HFE-7000. - Highlights: ► We studied the pool boiling performance of a thermosyphon with graphite foam evaporators of block and fin structures. ► FC-72 and HFE-7000 were used as the working fluids. ► The boiling heat transfer coefficients of the block structures are 1.2

  4. 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

  5. Chemical sputtering of graphite by H+ ions

    International Nuclear Information System (INIS)

    Busharov, N.P.; Gorbatov, E.A.; Gusev, V.M.; Guseva, M.I.; Martynenko, Y.V.

    1976-01-01

    In a study of the sputtering coefficient S for the sputtering of graphite by 10-keV H + ions as a function of the graphite temperature during the bombardment, it is found that at T> or =750degreeC the coefficient S is independent of the target temperature and has an anomalously high value, S=0.085 atom/ion. The high rate of sputtering of graphite by atomic hydrogen ions is shown to be due to chemical sputtering of the graphite, resulting primarily in the formation of CH 4 molecules. At T=1100degreeC, S falls off by a factor of about 3. A model for the chemical sputtering of graphite is proposed

  6. Graphite selection for the FMIT test cell

    International Nuclear Information System (INIS)

    Morgan, W.C.

    1982-06-01

    This document provides the basis for procuring a grade of graphite, at minimum cost, having minimum dimensional changes at low irradiation temperatures (nominal range 90 to 140 0 C). In light of those constraints, the author concludes that the most feasible approach is to attempt to reproduce a grade of graphite (TSGBF) which has exhibited a high degree of dimensional stability during low-temperature irradiations and on which irradiation-induced changes in other physical properties have been measured. The effects of differences in raw materials, especially coke morphology, and processing conditions, primarily graphitization temperture are briefly reviewed in terms of the practicality of producing a new grade of graphite with physical properties and irradiation-induced changes which would be very similar to those of TSGBF graphite. The production history and physical properties of TSGBF are also reviewed; no attempt is made, to project changes in dimensions or physical properties under the projected irradiation conditions

  7. Fuel management of HTR-10

    International Nuclear Information System (INIS)

    Wu Zongxin; Jing Xingqing

    2001-01-01

    The 10 MW high temperature cooled reactor (HTR-10) built in Tsinghua University is a pebble bed type of HTGR. The continuous recharge and multiple-pass of spherical fuel elements are used for fuel management. The initiative stage of core is composed of the mix of spherical fuel elements and graphite elements. The equilibrium stage of core is composed of identical spherical fuel elements. The fuel management during the transition from the initiative stage to the equilibrium stage is a key issue for HTR-10 physical design. A fuel management strategy is proposed based on self-adjustment of core reactivity. The neutron physical code is used to simulate the process of fuel management. The results show that the graphite elements, the recharging fuel elements below the burn-up allowance, and the discharging fuel elements over the burn-up allowance could be identified by burn-up measurement. The maximum of burn-up fuel elements could be controlled below the burn-up limit

  8. AGC-2 Graphite Preirradiation Data Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    William Windes; W. David Swank; David Rohrbaugh; Joseph Lord

    2013-08-01

    This report described the specimen loading order and documents all pre-irradiation examination material property measurement data for the graphite specimens contained within the second Advanced Graphite Capsule (AGC-2) irradiation capsule. The AGC-2 capsule is the second in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens allowing quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades to be generated which will ascertain the in-service behavior of the graphite for pebble bed and prismatic Very High Temperature Reactor (VHTR) designs. Similar to the AGC-1 specimen pre-irradiation examination report, material property tests were conducted on specimens from 18 nuclear graphite types but on an increased number of specimens (512) prior to loading into the AGC-2 irradiation assembly. All AGC-2 specimen testing was conducted at Idaho National Laboratory (INL) from October 2009 to August 2010. This report also details the specimen loading methodology for the graphite specimens inside the AGC-2 irradiation capsule. The AGC-2 capsule design requires “matched pair” creep specimens that have similar dose levels above and below the neutron flux profile mid-plane to provide similar specimens with and without an applied load. This document utilized the neutron flux profile calculated for the AGC-2 capsule design, the capsule dimensions, and the size (length) of the selected graphite and silicon carbide samples to create a stacking order that can produce “matched pairs” of graphite samples above and below the AGC-2 capsule elevation mid-point to provide specimens with similar neutron dose levels.

  9. Modeling Fission Product Sorption in Graphite Structures

    International Nuclear Information System (INIS)

    Szlufarska, Izabela; Morgan, Dane; Allen, Todd

    2013-01-01

    The goal of this project is to determine changes in adsorption and desorption of fission products to/from nuclear-grade graphite in response to a changing chemical environment. First, the project team will employ principle calculations and thermodynamic analysis to predict stability of fission products on graphite in the presence of structural defects commonly observed in very high-temperature reactor (VHTR) graphites. Desorption rates will be determined as a function of partial pressure of oxygen and iodine, relative humidity, and temperature. They will then carry out experimental characterization to determine the statistical distribution of structural features. This structural information will yield distributions of binding sites to be used as an input for a sorption model. Sorption isotherms calculated under this project will contribute to understanding of the physical bases of the source terms that are used in higher-level codes that model fission product transport and retention in graphite. The project will include the following tasks: Perform structural characterization of the VHTR graphite to determine crystallographic phases, defect structures and their distribution, volume fraction of coke, and amount of sp2 versus sp3 bonding. This information will be used as guidance for ab initio modeling and as input for sorptivity models; Perform ab initio calculations of binding energies to determine stability of fission products on the different sorption sites present in nuclear graphite microstructures. The project will use density functional theory (DFT) methods to calculate binding energies in vacuum and in oxidizing environments. The team will also calculate stability of iodine complexes with fission products on graphite sorption sites; Model graphite sorption isotherms to quantify concentration of fission products in graphite. The binding energies will be combined with a Langmuir isotherm statistical model to predict the sorbed concentration of fission products

  10. Ti-doped isotropic graphite: A promising armour material for plasma-facing components

    Science.gov (United States)

    García-Rosales, C.; López-Galilea, I.; Ordás, N.; Adelhelm, C.; Balden, M.; Pintsuk, G.; Grattarola, M.; Gualco, C.

    2009-04-01

    Finely dispersed Ti-doped isotropic graphites with 4 at.% Ti have been manufactured using synthetic mesophase pitch 'AR' as raw material. These new materials show a thermal conductivity at room temperature of ˜200 W/mK and flexural strength close to 100 MPa. Measurement of the total erosion yield by deuterium bombardment at ion energies and sample temperatures for which pure carbon shows maximum values, resulted in a reduction of at least a factor of 4, mainly due to dopant enrichment at the surface caused by preferential erosion of carbon. In addition, ITER relevant thermal shock loads were applied with an energetic electron beam at the JUDITH facility. The results demonstrated a significantly improved performance of Ti-doped graphite compared to pure graphite. Finally, Ti-doped graphite was successfully brazed to a CuCrZr block using a Mo interlayer. These results let assume that Ti-doped graphite can be a promising armour material for divertor plasma-facing components.

  11. Ti-doped isotropic graphite: A promising armour material for plasma-facing components

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Rosales, C. [CEIT and Tecnun (University of Navarra), Paseo de Manuel Lardizabal, 15, E-20018 San Sebastian (Spain)], E-mail: cgrosales@ceit.es; Lopez-Galilea, I.; Ordas, N. [CEIT and Tecnun (University of Navarra), Paseo de Manuel Lardizabal, 15, E-20018 San Sebastian (Spain); Adelhelm, C.; Balden, M. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, D-85748 Garching (Germany); Pintsuk, G. [Forschungszentrum Juelich GmbH, EURATOM Association, D-52425 Juelich (Germany); Grattarola, M.; Gualco, C. [Ansaldo Ricerche S.p.A., I-16152 Genoa (Italy)

    2009-04-30

    Finely dispersed Ti-doped isotropic graphites with 4 at.% Ti have been manufactured using synthetic mesophase pitch 'AR' as raw material. These new materials show a thermal conductivity at room temperature of {approx}200 W/mK and flexural strength close to 100 MPa. Measurement of the total erosion yield by deuterium bombardment at ion energies and sample temperatures for which pure carbon shows maximum values, resulted in a reduction of at least a factor of 4, mainly due to dopant enrichment at the surface caused by preferential erosion of carbon. In addition, ITER relevant thermal shock loads were applied with an energetic electron beam at the JUDITH facility. The results demonstrated a significantly improved performance of Ti-doped graphite compared to pure graphite. Finally, Ti-doped graphite was successfully brazed to a CuCrZr block using a Mo interlayer. These results let assume that Ti-doped graphite can be a promising armour material for divertor plasma-facing components.

  12. Ti-doped isotropic graphite: A promising armour material for plasma-facing components

    International Nuclear Information System (INIS)

    Garcia-Rosales, C.; Lopez-Galilea, I.; Ordas, N.; Adelhelm, C.; Balden, M.; Pintsuk, G.; Grattarola, M.; Gualco, C.

    2009-01-01

    Finely dispersed Ti-doped isotropic graphites with 4 at.% Ti have been manufactured using synthetic mesophase pitch 'AR' as raw material. These new materials show a thermal conductivity at room temperature of ∼200 W/mK and flexural strength close to 100 MPa. Measurement of the total erosion yield by deuterium bombardment at ion energies and sample temperatures for which pure carbon shows maximum values, resulted in a reduction of at least a factor of 4, mainly due to dopant enrichment at the surface caused by preferential erosion of carbon. In addition, ITER relevant thermal shock loads were applied with an energetic electron beam at the JUDITH facility. The results demonstrated a significantly improved performance of Ti-doped graphite compared to pure graphite. Finally, Ti-doped graphite was successfully brazed to a CuCrZr block using a Mo interlayer. These results let assume that Ti-doped graphite can be a promising armour material for divertor plasma-facing components.

  13. 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

  14. 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.

  15. Characterization of fresh and irradiated domestic nuclear graphite; Karakterizacija neozracenog i ozracenog domaceg nuklearnog grafita

    Energy Technology Data Exchange (ETDEWEB)

    Marinkovic, S; Suznjevic, C; Bogdanovic, R; Gasic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    This report shows results of testing the quality of domestic impregnated graphite IGSP-05, and irradiated domestic graphite IGSP-01 as well as the new methos of characterization based on graphite oxidation by liquid agent. Systematic measurement of domestic impregnated graphite enabled conclusions related to its quality and further improvement. Domestic graphite is relatively well graphitized and its properties are approaching standard nuclear graphite, although it still shows some deficiencies. Important deficiencies are significant inhomogeneity and low density. The applied impregnation procedure did not improve significantly the quality of graphite, probably because the material which was impregnated had fine pores. To avoid this porosity it would be necessary to use material with higher granulation. Soot which was present in some blocks probably worsened the quality of graphite and caused dispersion of the obtained results. First tests of irradiated domestic graphite IGSP-01 showed that its behaviour does not differ from standard nuclear graphite in case of low doses. It is necessary to test its properties in case of higher neutron doses before drawing final conclusions. The new method of graphite oxidation by the N{sub 2}SO{sub 4} - Ag{sub 2}Cr{sub 2}O{sub 7} mixture which is highly sensitive on the existence of structural defects is based on detecting the oxidation rate of graphite by measuring the pressure of released CO{sub 2}. Application of the method for testing the domestic and American graphite showed that irradiation caused drastic changes of oxidation rates and similar behaviour of both graphite types. U ovom izvestaju su prikazani rezultati ispitivanja kvaliteta domaceg impregnisanog grafita IGSP-05, rezultati ispitivanja ozracenog domaceg grafita IGSP-01 i opisana je nova uvedena metoda karakterizacije zasnovana na oksidaciji grafita tecnim agensom. Sistematsko merenje osobina domaceg impregnisanog grafita je omogucilo donosenje zakljucaka o

  16. Backfilling of deposition tunnels, block alternative

    International Nuclear Information System (INIS)

    Keto, P.; Roennqvist, P.-E.

    2007-03-01

    This report presents a preliminary process description of backfilling the deposition tunnels with pre-compacted blocks consisting of a mixture of bentonite and ballast (30:70). The process was modified for the Finnish KBS-3V type repository assuming that the amount of spent fuel canisters disposed of yearly is 40. Backfilling blocks (400 x 300 x 300 mm) are prepared in a block production plant with a hydraulic press with an estimated production capacity of 840 blocks per day. Some of the blocks are modified further to fit the profile of the tunnel roof. Prior to the installation of the blocks, the deposition tunnel floor is levelled with a mixture of bentonite and ballast (15:85). The blocks are placed in the tunnel with a modified reach truck. Centrifugal pellet throwing equipment is used to fill the gap between the blocks and the rock surface with bentonite pellets. Based on a preliminary assessment, the average dry density achieved with block backfill is sufficient to fulfil the criteria set for the backfill in order to ensure long-term safety and radiation protection. However, there are uncertainties concerning saturation, homogenisation, erosion, piping and self-healing of the block backfill that need to be studied further with laboratory and field tests. In addition, development efforts and testing concerning block manufacturing and installation are required to verify the technical feasibility of the concept. (orig.)

  17. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

    Lotts, A.L.; Homan, F.J.; Balthesen, E.; Turner, R.F.

    1977-01-01

    Significant advances have occurred in the development of HTGR fuel and fuel cycle. These accomplishments permit a wide choice of fuel designs, reactor concepts, and fuel cycles. Fuels capable of providing helium outlet temperatures of 750 0 C are available, and fuels capable of 1000 0 C outlet temperatures may be expected from extension of present technology. Fuels have been developed for two basic HTGR designs, one using a spherical (pebble bed) element and the other a prismatic element. Within each concept a number of variations of geometry, fuel composition, and structural materials are permitted. Potential fuel cycles include both low-enriched and high-enriched Th- 235 U, recycle Th- 233 U, and Th-Pu or U-Pu cycles. This flexibility offered by the HTGR is of great practical benefit considering the rapidly changing economics of power production. The inflation of ore prices has increased optimum conversion ratios, and increased the necessity of fuel recycle at an early date. Fuel element makeup is very similar for prismatic and spherical designs. Both use spherical fissile and fertile particles coated with combinations of pyrolytic carbon and silicon carbide. Both use carbonaceous binder materials, and graphite as the structural material. Weak-acid resin (WAR) UO 2 -UC 2 fissile fuels and sol-gel-derived ThO 2 fertile fuels have been selected for the Th- 233 U cycle in the prismatic design. Sol-gel-derived UO 2 UC 2 is the reference fissile fuel for the low-enriched pebble bed design. Both the United States and Federal Republic of Germany are developing technology for fuel cycle operations including fabrication, reprocessing, refabrication, and waste handling. Feasibility of basic processes has been established and designs developed for full-scale equipment. Fuel and fuel cycle technology provide the basis for a broad range of applications of the HTGR. Extension of the fuels to higher operating temperatures and development and commercial demonstration of fuel

  18. Fuel performance in water storage

    International Nuclear Information System (INIS)

    Hoskins, A.P.; Scott, J.G.; Shelton-Davis, C.V.; McDannel, G.E.

    1993-11-01

    Westinghouse Idaho Nuclear Company operates the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering Laboratory (INEL) for the Department of Energy (DOE). A variety of different types of fuels have been stored there since the 1950's prior to reprocessing for uranium recovery. In April of 1992, the DOE decided to end fuel reprocessing, changing the mission at ICPP. Fuel integrity in storage is now viewed as long term until final disposition is defined and implemented. Thus, the condition of fuel and storage equipment is being closely monitored and evaluated to ensure continued safe storage. There are four main areas of fuel storage at ICPP: an original underwater storage facility (CPP-603), a modern underwater storage facility (CPP-666), and two dry fuel storage facilities. The fuels in storage are from the US Navy, DOE (and its predecessors the Energy Research and Development Administration and the Atomic Energy Commission), and other research programs. Fuel matrices include uranium oxide, hydride, carbide, metal, and alloy fuels. In the underwater storage basins, fuels are clad with stainless steel, zirconium, and aluminum. Also included in the basin inventory is canned scrap material. The dry fuel storage contains primarily graphite and aluminum type fuels. A total of 55 different fuel types are currently stored at the Idaho Chemical Processing Plant. The corrosion resistance of the barrier material is of primary concern in evaluating the integrity of the fuel in long term water storage. The barrier material is either the fuel cladding (if not canned) or the can material

  19. Low cost sic coated erosion resistant graphite

    International Nuclear Information System (INIS)

    Zafar, M.F.; Nicholls, J.R.

    2007-01-01

    The development of materials with unique and improved properties using low cost processes is essential to increase performance and reduce cost of the solid rocket motors. Specifically advancements are needed for boost phase nozzle. As these motors operate at very high pressure and temperatures, the nozzle must survive high thermal stresses with minimal erosion to maintain performance. Currently three material choices are being exploited; which are refractory metals, graphite and carbon-carbon composites. Of these three materials graphite is the most attractive choice because of its low cost, light weight, and easy forming. However graphite is prone to erosion, both chemical and mechanical, which may affect the ballistic conditions and mechanical properties of the nozzle. To minimize this erosion high density graphite is usually preferred; which is again very expensive. Another technique used to minimize the erosion is Pyrolytic Graphite (PG) coating inside the nozzle. However PG coating is prone to cracking and spallation along with very cumbersome deposition process. Another possible methodology to avoid this erosion is to convert the inside surface of the rocket nozzle to Silicon Carbide (SiC), which is very erosion resistant and have much better thermal stability compared to graphite and even PG. Due to its functionally gradient nature such a layer will be very adherent and resistant to spallation. The current research is focused on synthesizing, characterizing and oxidation testing of such a converted SiC layer on commercial grade graphite. (author)

  20. 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.

  1. Biaxial testing for nuclear grade graphite by ball on three balls assessment

    International Nuclear Information System (INIS)

    Mohd Reusmaazran Yusof; Yusof Abdullah

    2012-01-01

    Nuclear grade (high-purity) graphite for fuel element and moderator material in Advanced Gas Cooling Reactors (AGR) displays large scatter in strength and a non-linear stress-strain response from the damage accumulation. These responses can be characterized as quasi-brittle behaviour. Current assessments of fracture in core graphite components are based on the linear elastic approximation and thus represent a major assumption. The quasi-brittle behaviour gives challenge to assess the real nuclear graphite component. The selected test method would help to bridge the gap between microscale to macro-scale in real reactor component. The small scale tests presented here can contribute some statistical data to manifests the failure in real component. The evaluation and choice of different solution design of biaxial test will be discussed in this paper. The ball on-three ball test method was used for assessment test follows by numerous of analytical method. The results shown that biaxial strength of the EY9 grade graphite depends on the method used for evaluation. Some of the analytical methods use to calculate biaxial strength were found not to be valid and therefore should not be used to assess the mechanical properties of nuclear graphite. (author)

  2. Development and testing of nuclear graphite for the German pebble-bed high temperature reactor

    International Nuclear Information System (INIS)

    Haag, G.; Delle, W.; Nickel, H.; Theymann, W.; Wilhelmi, G.

    1987-01-01

    Several types of high temperature reactors have been developed in the Federal Republic of Germany. They are all based on spherical fuel elements being surrounded by graphite as reflector material. As an example, HTR-500 developed by the Hochtemperatur Reaktorbau GmbH is shown. The core consists of the top reflector, the side reflector with inner and outer parts, the bottom reflector and the core support columns. The most serious problem with respect to fast neutron radiation damage had to be solved for the materials of those parts near the pebble bed. Regarding the temperature profile in the core, the top reflector is at 300 deg C, and as cooling gas flows from the top downward, the temperature of the inner side reflector rises to about 700 deg C at the bottom. Fortunately, the highest fast neutron load accumulated during the life time of a reactor corresponds to the lowest temperature. This makes graphite components easier to survive neutron exposure without being mechanically damaged, although the maximum fast neutron fluence is as high as 4 x 10 22 /cm 2 at about 400 deg C. HTR graphite components are divided into four classes according to loading. The raw materials for nuclear graphite, the development of pitch coke nuclear graphite, the irradiation behavior of ATR-2E and ASR-IRS and others are reported. (Kako, I.)

  3. Assessment of fuel concepts

    International Nuclear Information System (INIS)

    Bailey, W.J.; Barner, J.O.

    1978-01-01

    The relative merits of various LWR UO 2 fuel concepts with the potential for improved power-ramping capability were qualitatively assessed. In the evaluation, it was determined that of the various concepts being considered, those that presently possess an adequately developed experience base include annular pellets, cladding coated with graphite on the inner surface, and packed-particle fuel. Therefore, these were selected for initial evaluation as part of the Fuel Performance Improvement Program. For this program, graphite-coated cladding is being used in conjunction with annular pellet fuel as one of the concepts with the anticipation of gaining the advantage of the combined improvements. The report discusses the following: the criteria used to evaluate the candidate fuel concepts; a comparison of the concepts selected for irradiation with the criteria, including a general description of their experience bases; and a general discussion of other candidate concepts, including identifying those which may be considered for out-of-reactor evaluation as part of this program, those for which the results of other programs will be monitored, and those which have been deleted from further consideration at this time

  4. Experimental study of some mounting brackets to support fuel elements

    International Nuclear Information System (INIS)

    Aubert, M.; Poglia, S.; Roche, R.

    1958-09-01

    In an atomic pile with vertical channels, fuel elements are stacked on one another. According to a possible assembly, fuel element can be contained by a graphite sleeve and be supported by a mounting bracket in this sleeve. Sleeves are then stacked on one another. The authors report the investigation of different designs for these mounting brackets. They describe their mechanical role and their mechanical, aerodynamic, neutronic and test conditions. They report tests performed on brackets made in graphite and on brackets made in stainless steel and graphite, and discuss the obtained results

  5. Investigation of Hydrogen and Nitrogen Content in Compacted Graphite Iron Production

    OpenAIRE

    Siafakas, Dimitrios

    2013-01-01

    The aim of this research, part of a wider program called SPOFIC, is to investigate how the casting procedure affects the concentration of hydrogen and nitrogen gases in Compacted Graphite Iron used for the production of truck cylinder blocks. Hydris equipment was used for the Hydrogen measurements and the Optical Emission Spectroscopy and combustion analysis methods were used for the nitrogen measurements. The experiment was performed in one of the cooperating foundries. It was found that Hyd...

  6. 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

  7. Synthesis of soluble graphite and graphene.

    Science.gov (United States)

    Kelly, K F; Billups, W E

    2013-01-15

    Because of graphene's anticipated applications in electronics and its thermal, mechanical, and optical properties, many scientists and engineers are interested in this material. Graphene is an isolated layer of the π-stacked hexagonal allotrope of carbon known as graphite. The interlayer cohesive energy of graphite, or exfoliation energy, that results from van der Waals attractions over the interlayer spacing distance of 3.34 Å (61 meV/C atom) is many times weaker than the intralayer covalent bonding. Since graphene itself does not occur naturally, scientists and engineers are still learning how to isolate and manipulate individual layers of graphene. Some researchers have relied on the physical separation of the sheets, a process that can sometimes be as simple as peeling of sheets from crystalline graphite using Scotch tape. Other researchers have taken an ensemble approach, where they exploit the chemical conversion of graphite to the individual layers. The typical intermediary state is graphite oxide, which is often produced using strong oxidants under acidic conditions. Structurally, researchers hypothesize that acidic functional groups functionalize the oxidized material at the edges and a network of epoxy groups cover the sp(2)-bonded carbon network. The exfoliated material formed under these conditions can be used to form dispersions that are usually unstable. However, more importantly, irreversible defects form in the basal plane during oxidation and remain even after reduction of graphite oxide back to graphene-like material. As part of our interest in the dissolution of carbon nanomaterials, we have explored the derivatization of graphite following the same procedures that preserve the sp(2) bonding and the associated unique physical and electronic properties in the chemical processing of single-walled carbon nanotubes. In this Account, we describe efficient routes to exfoliate graphite either into graphitic nanoparticles or into graphene without

  8. Adsorption of lead over graphite oxide.

    Science.gov (United States)

    Olanipekun, Opeyemi; Oyefusi, Adebola; Neelgund, Gururaj M; Oki, Aderemi

    2014-01-24

    The adsorption efficiency and kinetics of removal of lead in presence of graphite oxide (GO) was determined using the Atomic Absorption Spectrophotometer (AAS). The GO was prepared by the chemical oxidation of graphite and characterized using FTIR, SEM, TGA and XRD. The adsorption efficiency of GO for the solution containing 50, 100 and 150 ppm of Pb(2+) was found to be 98%, 91% and 71% respectively. The adsorption ability of GO was found to be higher than graphite. Therefore, the oxidation of activated carbon in removal of heavy metals may be a viable option to reduce pollution in portable water. Published by Elsevier B.V.

  9. Interface structure between tetraglyme and graphite

    Science.gov (United States)

    Minato, Taketoshi; Araki, Yuki; Umeda, Kenichi; Yamanaka, Toshiro; Okazaki, Ken-ichi; Onishi, Hiroshi; Abe, Takeshi; Ogumi, Zempachi

    2017-09-01

    Clarification of the details of the interface structure between liquids and solids is crucial for understanding the fundamental processes of physical functions. Herein, we investigate the structure of the interface between tetraglyme and graphite and propose a model for the interface structure based on the observation of frequency-modulation atomic force microscopy in liquids. The ordering and distorted adsorption of tetraglyme on graphite were observed. It is found that tetraglyme stably adsorbs on graphite. Density functional theory calculations supported the adsorption structure. In the liquid phase, there is a layered structure of the molecular distribution with an average distance of 0.60 nm between layers.

  10. 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

  11. The fuel of nuclear reactors

    International Nuclear Information System (INIS)

    1995-03-01

    This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

  12. Ultrasound guided supraclavicular block.

    LENUS (Irish Health Repository)

    Hanumanthaiah, Deepak

    2013-09-01

    Ultrasound guided regional anaesthesia is becoming increasingly popular. The supraclavicular block has been transformed by ultrasound guidance into a potentially safe superficial block. We reviewed the techniques of performing supraclavicular block with special focus on ultrasound guidance.

  13. Noncovalently functionalized graphitic mesoporous carbon as a stable support of Pt nanoparticles for oxygen reduction

    Energy Technology Data Exchange (ETDEWEB)

    Shao, Yuyan; Zhang, Sheng; Kou, Rong; Wang, Chongmin; Viswanathan, Vilayanur; Liu, Jun; Wang, Yong; Lin, Yuehe [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Wang, Xiqing; Dai, Sheng [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2010-04-02

    We report a durable electrocatalyst support, highly graphitized mesoporous carbon (GMPC), for oxygen reduction in polymer electrolyte membrane (PEM) fuel cells. GMPC is prepared through graphitizing the self-assembled soft-template mesoporous carbon (MPC) under high temperature. Heat-treatment at 2800 C greatly improves the degree of graphitization while most of the mesoporous structures and the specific surface area of MPC are retained. GMPC is then noncovalently functionalized with poly(diallyldimethylammonium chloride) (PDDA) and loaded with Pt nanoparticles by reducing Pt precursor (H{sub 2}PtCl{sub 6}) in ethylene glycol. Pt nanoparticles of {proportional_to}3.0 nm in diameter are uniformly dispersed on GMPC. Compared to Pt supported on Vulcan XC-72 carbon black (Pt/XC-72), Pt/GMPC exhibits a higher mass activity towards oxygen reduction reaction (ORR) and the mass activity retention (in percentage) is improved by a factor of {proportional_to}2 after 44 h accelerated degradation test under the potential step (1.4-0.85 V) electrochemical stressing condition which focuses on support corrosion. The enhanced activity and durability of Pt/GMPC are attributed to the graphitic structure of GMPC which is more resistant to corrosion. These findings demonstrate that GMPC is a promising oxygen reduction electrocatalyst support for PEM fuel cells. The approach reported in this work provides a facile, eco-friendly promising strategy for synthesizing stable metal nanoparticles on hydrophobic support materials. (author)

  14. Superhydrophilic graphite surfaces and water-dispersible graphite colloids by electrochemical exfoliation

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yueh-Feng [Department of Chemical and Materials Engineering, National Central University, Jhongli, 320 Taiwan (China); Chen, Shih-Ming; Lai, Wei-Hao [Materials and Chemical Research Laboratories, Industrial Technology Research Institute, Chutung, Hsinchu, 31040 Taiwan (China); Sheng, Yu-Jane [Department of Chemical Engineering, National Taiwan University, Taipei, 106 Taiwan (China); Tsao, Heng-Kwong [Department of Chemical and Materials Engineering, Department of Physics, National Central University, Jhongli, 320 Taiwan (China)

    2013-08-14

    Superhydrophilic graphite surfaces and water-dispersible graphite colloids are obtained by electrochemical exfoliation with hydrophobic graphite electrodes. Such counterintuitive characteristics are caused by partial oxidation and investigated by examining both graphite electrodes and exfoliated particles after electrolysis. The extent of surface oxidation can be explored through contact angle measurement, scanning electron microscope, electrical sheet resistance, x-ray photoelectron spectroscopy, zeta-potential analyzer, thermogravimetric analysis, UV-visible, and Raman spectroscopy. The degree of wettability of the graphite anode can be altered by the electrolytic current and time. The water contact angle declines generally with increasing the electrolytic current or time. After a sufficient time, the graphite anode becomes superhydrophilic and its hydrophobicity can be recovered by peeling with adhesive tape. This consequence reveals that the anodic graphite is oxidized by oxygen bubbles but the oxidation just occurs at the outer layers of the graphite sheet. Moreover, the characteristics of oxidation revealed by UV peak shift, peak ratio between D and G bands, and negative zeta-potential indicate the presence of graphite oxide on the outer shell of the exfoliated colloids. However, thermogravimetric analysis for the extent of decomposition of oxygen functional groups verifies that the amount of oxygen groups is significantly less than that of graphite oxide prepared via Hummer method. The structure of this partially oxidized graphite may consist of a graphite core covered with an oxidized shell. The properties of the exfoliated colloids are also influenced by pH of the electrolytic solution. As pH is increased, the extent of oxidation descends and the thickness of oxidized shell decreases. Those results reveal that the degree of oxidation of exfoliated nanoparticles can be manipulated simply by controlling pH.

  15. Performance of AC/graphite capacitors at high weight ratios of AC/graphite

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Hongyu [IM and T Ltd., Advanced Research Center, Saga University, 1341 Yoga-machi, Saga 840-0047 (Japan); Yoshio, Masaki [Advanced Research Center, Department of Applied Chemistry, Saga University, 1341 Yoga-machi, Saga 840-0047 (Japan)

    2008-03-01

    The effect of negative to positive electrode materials' weight ratio on the electrochemical performance of both activated carbon (AC)/AC and AC/graphite capacitors has been investigated, especially in the terms of capacity and cycle-ability. The limited capacity charge mode has been proposed to improve the cycle performance of AC/graphite capacitors at high weight ratios of AC/graphite. (author)

  16. Fabrication technology of spherical fuel element for HTR-10

    International Nuclear Information System (INIS)

    He Jun; Zou Yanwen; Liang Tongxiang; Qiu Xueliang

    2002-01-01

    R and D on the fabrication technology of the spherical fuel elements for the 10 MW HTR Test Module (HTR-10) began from 1986. Cold quasi-isostatic molding with a silicon rubber die is used for manufacturing the spherical fuel elements.The fabrication technology and the graphite matrix materials were investigated and optimized. Twenty five batches of fuel elements, about 11000 of the fuel elements, have been produced. The cold properties of the graphite matrix materials satisfied the design specifications. The mean free uranium fraction of 25 batches was 5 x 10 -5

  17. Improved techniques for the ultrasonic characterization of graphite

    International Nuclear Information System (INIS)

    Cook, K.V.; Simpson, W.A.

    1981-01-01

    Near-conventional pulse-echo flaw detection techniques can be successfully applied to graphite material. If the specimen configuration (i.e., length-to-diameter ratio) is conducive to the choice of ultrasonic test parameters dictated by the material screening tests, then only the test sensitivity needs to be established. Conventional test block approaches to calibration work well; however, uniform homogeneity of the specimens to be tested must be assumed (depending on the graphite, this may be an invalid assumption). Sensitivities that we have demonstrated typically detect 0.5- to 0.75-mm (0.020 to 0.030-in.) reflectors at depths up to 102 mm (4 in.) for GRAPHNOL (drilled holes, not flat bottom). Other materials may dictate inspection for much larger discontinuities. The least sensitive tests performed to date (using the storage oscilloscope approach) required flat-bottom holes of 6.35 mm (0.250 in.) in diameter for calibration. This relatively insensitive test was necessary because of the billet length and material characteristics

  18. Management of radioactive waste in nuclear power: handling of irradiated graphite from water-cooled graphite reactors

    International Nuclear Information System (INIS)

    Anfimov, S.S.

    2001-01-01

    In this paper an radioactive waste processing of graphite from graphite moderated nuclear reactors at its decommissioning is discussed. Methods of processing of irradiated graphite are presented. It can be concluded that advanced methods for graphite radioactive waste handling are available nowadays. Implementation of these methods will allow to enhance environmental safety of nuclear power that will benefit its progress in the future

  19. Ab initio and Molecular Dynamic models of displacement damage in crystalline and turbostratic graphite

    Science.gov (United States)

    McKenna, Alice

    One of the functions of graphite is as a moderator in several nuclear reactor designs, including the Advanced Gas-cooled Reactor (AGR). In the reactor graphite is used to thermalise the neutrons produced in the fission reaction thus allowing a self-sustained reaction to occur. The graphite blocks, acting as the moderator, are constantly irradiated and consequently suffer damage. This thesis examines the types of damage caused using molecular dynamic (MD) simulations and ab intio calculations. Neutron damage starts with a primary knock-on atom (PKA), which is travelling so fast that it creates damage through electronic and thermal excitation (this is addressed with thermal spike simulations). When the PKA has lost energy the subsequent cascade is based on ballistic atomic displacement. These two types of simulations were performed on single crystal graphite and other carbon structures such as diamond and amorphous carbon as a comparison. The thermal spike in single crystal graphite produced results which varied from no defects to a small number of permanent defects in the structure. It is only at the high energy range that more damage is seen but these energies are less likely to occur in the nuclear reactor. The thermal spike does not create damage but it is possible that it can heal damaged sections of the graphite, which can be demonstrated with the motion of the defects when a thermal spike is applied. The cascade simulations create more damage than the thermal spike even though less energy is applied to the system. A new damage function is found with a threshold region that varies with the square root of energy in excess of the energy threshold. This is further broken down in to contributions from primary and subsequent knock-on atoms. The threshold displacement energy (TDE) is found to be Ed=25eV at 300K. In both these types of simulation graphite acts very differently to the other carbon structures. There are two types of polycrystalline graphite structures

  20. 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)

  1. Synthesis of graphene nanoplatelets from peroxosulfate graphite intercalation compounds

    OpenAIRE

    MELEZHYK A.V.; TKACHEV A.G.

    2014-01-01

    Ultrasonic exfoliation of expanded graphite compound obtained by cold expansion of graphite intercalated with peroxodisulfuric acid was shown to allow the creation of graphene nanoplatelets with thickness of about 5-10 nm. The resulting graphene material contained surface oxide groups. The expanded graphite intercalation compound was exfoliated by ultrasound much easier than thermally expanded graphite. A mechanism for the cleavage of graphite to graphene nanoplatelets is proposed. It include...

  2. 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

  3. 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

  4. 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)

  5. 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.)

  6. Immobilization of Rocky Flats Graphite Fines Residue

    International Nuclear Information System (INIS)

    Rudisill, T.S.

    1999-01-01

    The development of the immobilization process for graphite fines has proceeded through a series of experimental programs. The experimental procedures and results from each series of experiments are discussed in this report

  7. Optical motion control of maglev graphite.

    Science.gov (United States)

    Kobayashi, Masayuki; Abe, Jiro

    2012-12-26

    Graphite has been known as a typical diamagnetic material and can be levitated in the strong magnetic field. Here we show that the magnetically levitating pyrolytic graphite can be moved in the arbitrary place by simple photoirradiation. It is notable that the optical motion control system described in this paper requires only NdFeB permanent magnets and light source. The optical movement is driven by photothermally induced changes in the magnetic susceptibility of the graphite. Moreover, we demonstrate that light energy can be converted into rotational kinetic energy by means of the photothermal property. We find that the levitating graphite disk rotates at over 200 rpm under the sunlight, making it possible to develop a new class of light energy conversion system.

  8. 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

  9. Collective modes in superconducting rhombohedral graphite

    Energy Technology Data Exchange (ETDEWEB)

    Kauppila, Ville [O.V. Lounasmaa Laboratory, Aalto University (Finland); Hyart, Timo; Heikkilae, Tero [University of Jyvaeskylae (Finland)

    2015-07-01

    Recently it was realized that coupling particles with a Dirac dispersion (such as electrons in graphene) can lead to a topologically protected state with flat band dispersion. Such a state could support superconductivity with unusually high critical temperatures. Perhaps the most promising way to realize such coupling in real materials is in the surface of rhombohedrally stacked graphite. We consider collective excitations (i.e. the Higgs modes) in surface superconducting rhombohedral graphite. We find two amplitude and two phase modes corresponding to the two surfaces of the graphite where the superconductivity lives. We calculate the dispersion of these modes. We also derive the Ginzburg-Landau theory for this material. We show that in superconducting rhombohedral graphite, the collective modes, unlike in conventional BCS superconductors, give a large contribution to thermodynamic properties of the material.

  10. Large Scale Reduction of Graphite Oxide

    Data.gov (United States)

    National Aeronautics and Space Administration — This project seeks to develop an optical method to reduce graphite oxide into graphene efficiently and in larger formats than currently available. Current reduction...

  11. Analysis of picosecond pulsed laser melted graphite

    International Nuclear Information System (INIS)

    Steinbeck, J.; Braunstein, G.; Speck, J.; Dresselhaus, M.S.; Huang, C.Y.; Malvezzi, A.M.; Bloembergen, N.

    1986-01-01

    A Raman microprobe and high resolution TEM have been used to analyze the resolidified region of liquid carbon generated by picosecond pulse laser radiation. From the relative intensities of the zone center Raman-allowed mode for graphite at 1582 cm -1 and the disorder-induced mode at 1360 cm -1 , the average graphite crystallite size in the resolidified region is determined as a function of position. By comparison with Rutherford backscattering spectra and Raman spectra from nonosecond pulsed laser melting experiments, the disorder depth for picosecond pulsed laser melted graphite is determined as a function of irradiating energy density. Comparisons of TEM micrographs for nanosecond and picosecond pulsed laser melting experiments show that the structure of the laser disordered regions in graphite are similar and exhibit similar behavior with increasing laser pulse fluence

  12. Electrochemical oxidation of textile industry wastewater by graphite electrodes.

    Science.gov (United States)

    Bhatnagar, Rajendra; Joshi, Himanshu; Mall, Indra D; Srivastava, Vimal C

    2014-01-01

    In the present article, studies have been performed on the electrochemical (EC) oxidation of actual textile industry wastewater by graphite electrodes. Multi-response optimization of four independent parameters namely initial pH (pHo): 4-10, current density (j): 27.78-138.89 A/m(2), NaCl concentration (w): 0-2 g/L and electrolysis time (t): 10-130 min have been performed using Box-Behnken (BB) experimental design. It was aimed to simultaneously maximize the chemical oxygen demand (COD) and color removal efficiencies and minimize specific energy consumption using desirability function approach. Pareto analysis of variance (ANOVA) showed a high coefficient of determination value for COD (R(2) = 0.8418), color (R(2) = 0.7010) and specific energy (R(2) = 0.9125) between the experimental values and the predicted values by a second-order regression model. Maximum COD and color removal and minimum specific energy consumed was 90.78%, 96.27% and 23.58 kWh/kg COD removed, respectively, were observed at optimum conditions. The wastewater, sludge and scum obtained after treatment at optimum condition have been characterized by various techniques. UV-visible study showed that all azo bonds of the dyes present in the wastewater were totally broken and most of the aromatic rings were mineralized during EC oxidation with graphite electrode. Carbon balance showed that out of the total carbon eroded from the graphite electrodes, 27-29.2% goes to the scum, 71.1-73.3% goes into the sludge and rest goes to the treated wastewater. Thermogravimetric analysis showed that the generated sludge and scum can be dried and used as a fuel in the boilers/incinerators.

  13. Elastic properties of graphite and interstitial defects

    International Nuclear Information System (INIS)

    Ayasse, J.-B.

    1977-01-01

    The graphite elastic constants C 33 and C 44 , reflecting the interaction of the graphitic planes, were experimentally measured as a function of irradiation and temperature. A model of non-central strength atomic interaction was established to explain the experimental results obtained. This model is valid at zero temperature. The temperature dependence of the elastic properties was analyzed. The influence of the elastic property variations on the specific heat of the lattice at very low temperature was investigated [fr

  14. 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.)

  15. High temperature tests for graphite materials

    OpenAIRE

    Zhmurikov, Evgenij

    2015-01-01

    This study was performed within the framework of the EURISOL for facilities SPIRAL-II (GANIL, France) and SPES (LNL, Italy), and aims to investigate the anticipated strength properties of fine-grained graphite at elevated temperatures. It appears that the major parameters that affect to the lifetime of a graphite target of this IP are the temperature and heating time. High temperature tests were conducted to simulate the heating under the influence of a beam of heavy particles by passing thro...

  16. Thermal decomposition of cesium-ethylene-ternary graphite intercalation compounds

    International Nuclear Information System (INIS)

    Matsumoto, R.; Oishi, Y.; Arii, T.

    2010-01-01

    In this paper, the thermal decomposition of air-stable Cs-ethylene-ternary graphite intercalation compounds (GICs) is discussed. The air stability of Cs-GICs is improved remarkably after the absorption of ethylene into their interlayer nanospace, because the ethylene molecules oligomerize and block the movement of Cs atoms. In addition, the evaporation of Cs atoms from the Cs-ethylene-ternary GICs is observed above 400 o C under a N 2 atmosphere of 100 Pa by ion attachment mass spectrometry. Although the results indicate that Cs-ethylene-ternary GICs remain stable up to approximately 400 o C, their thermal stability is not very high as compared to that of Cs-GICs.

  17. Graphite Composite Panel Polishing Fixture

    Science.gov (United States)

    Hagopian, John; Strojny, Carl; Budinoff, Jason

    2011-01-01

    The use of high-strength, lightweight composites for the fixture is the novel feature of this innovation. The main advantage is the light weight and high stiffness-to-mass ratio relative to aluminum. Meter-class optics require support during the grinding/polishing process with large tools. The use of aluminum as a polishing fixture is standard, with pitch providing a compliant layer to allow support without deformation. Unfortunately, with meter-scale optics, a meter-scale fixture weighs over 120 lb (.55 kg) and may distort the optics being fabricated by loading the mirror and/or tool used in fabrication. The use of composite structures that are lightweight yet stiff allows standard techniques to be used while providing for a decrease in fixture weight by almost 70 percent. Mounts classically used to support large mirrors during fabrication are especially heavy and difficult to handle. The mount must be especially stiff to avoid deformation during the optical fabrication process, where a very large and heavy lap often can distort the mount and optic being fabricated. If the optic is placed on top of the lapping tool, the weight of the optic and the fixture can distort the lap. Fixtures to support the mirror during fabrication are often very large plates of aluminum, often 2 in. (.5 cm) or more in thickness and weight upwards of 150 lb (68 kg). With the addition of a backing material such as pitch and the mirror itself, the assembly can often weigh over 250 lb (.113 kg) for a meter-class optic. This innovation is the use of a lightweight graphite panel with an aluminum honeycomb core for use as the polishing fixture. These materials have been used in the aerospace industry as structural members due to their light weight and high stiffness. The grinding polishing fixture consists of the graphite composite panel, fittings, and fixtures to allow interface to the polishing machine, and introduction of pitch buttons to support the optic under fabrication. In its

  18. Behavior of LASL-made graphite, ZrC, and ZrC-containing coated particles in irradiation tests HT-28 and HT-29

    International Nuclear Information System (INIS)

    Reiswig, R.D.; Wagner, P.; Hollabaugh, C.M.; White, R.W.; O'Rourke, J.A.; Davidson, K.V.; Schell, D.H.

    1976-01-01

    Three types of materials, extruded graphite, hot-pressed ZrC, and particles with ZrC coatings, were irradiated in ORNL High Fluence Isotope Reactor Irradiation tests HT-28 and HT-29. The ZrC seemed unaffected. The graphite changed in dimensions, x-ray diffraction parameters, and thermal conductivity. The four types of coated particles tested all resisted the irradiation well, except one set of particles with double-graded C-ZrC-C coats. Overall, the results were considered encouraging for use of ZrC and extruded graphite fuel matrices. 16 fig

  19. 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)

  20. Device for reprocessing nuclear fuels

    International Nuclear Information System (INIS)

    Hatano, Mamoru.

    1981-01-01

    Purpose: To readily discharge a nuclear fuel by burning the nuclear fuel as it is without a pulverizing step and removing the graphite and other coated fuel particles. Constitution: An oxygen supply pipe is connected to the lower po