WorldWideScience

Sample records for graphite target operating

  1. 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

  2. Graphite target for the spiral project

    International Nuclear Information System (INIS)

    Putaux, J.C.; Ducourtieux, M.; Ferro, A.; Foury, P.; Kotfila, L.; Mueller, A.C.; Obert, J.; Pauwels, N.; Potier, J.C.; Proust, J.; Loiselet, M.

    1996-01-01

    A study of the thermal and physical properties of graphite targets for the SPIRAL project is presented. The main objective is to develop an optimized set-up both mechanically and thermally resistant, presenting good release properties (hot targets with thin slices). The results of irradiation tests concerning the mechanical and thermal resistance of the first prototype of SPIRAL target with conical geometry are presented. The micro-structural properties of the graphite target is also studied, in order to check that the release properties are not deteriorated by the irradiation. Finally, the results concerning the latest pilot target internally heated by an electrical current are shown. (author)

  3. 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.

  4. Preparation of graphite targets for radiocarbon dating by tandem accelerator mass spectrometer (TAMS)

    International Nuclear Information System (INIS)

    Lowe, D.C.

    1984-01-01

    The introduction of TAMS has exciting implications for radiocarbon dating but improved sample preparation methods are needed. This paper describes a promising method for the conversion of a few milligrams of wood or charcoal into graphite targets for use in a caesium sputter ion source. Targets containing a large proportion of G-type graphite produced large C - currents, but those containing a high proportion of turbostatic Tn graphite were unsatisfactory; the type of graphite in the target is clearly of significance. (author)

  5. 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

  6. Process for the fabrication of aluminum metallized pyrolytic graphite sputtering targets

    Science.gov (United States)

    Makowiecki, Daniel M.; Ramsey, Philip B.; Juntz, Robert S.

    1995-01-01

    An improved method for fabricating pyrolytic graphite sputtering targets with superior heat transfer ability, longer life, and maximum energy transmission. Anisotropic pyrolytic graphite is contoured and/or segmented to match the erosion profile of the sputter target and then oriented such that the graphite's high thermal conductivity planes are in maximum contact with a thermally conductive metal backing. The graphite contact surface is metallized, using high rate physical vapor deposition (HRPVD), with an aluminum coating and the thermally conductive metal backing is joined to the metallized graphite target by one of four low-temperature bonding methods; liquid-metal casting, powder metallurgy compaction, eutectic brazing, and laser welding.

  7. Post Irradiation Examination Results of the NT-02 Graphite Fins NUMI Target

    Energy Technology Data Exchange (ETDEWEB)

    Ammigan, K. [Fermilab; Hurh, P. [Fermilab; Sidorov, V. [Fermilab; Zwaska, R. [Fermilab; Asner, D. M. [PNL, Richland; Casella, Casella,A.M [PNL, Richland; Edwards, D. J. [PNL, Richland; Schemer-Kohrn, A. L. [PNL, Richland; Senor, D. J. [PNL, Richland

    2017-02-10

    The NT-02 neutrino target in the NuMI beamline at Fermilab is a 95 cm long target made up of segmented graphite fins. It is the longest running NuMI target, which operated with a 120 GeV proton beam with maximum power of 340 kW, and saw an integrated total proton on target of 6.1 1020. Over the last half of its life, gradual degradation of neutrino yield was observed until the target was replaced. The probable causes for the target performance degradation are attributed to radiation damage, possibly including cracking caused by reduction in thermal shock resistance, as well as potential localized oxidation in the heated region of the target. Understanding the long-termstructural response of target materials exposed to proton irradiation is critical as future proton accelerator sources are becoming increasingly more powerful. As a result, an autopsy of the target was carried out to facilitate post-irradiation examination of selected graphite fins. Advanced microstructural imaging and surface elemental analysis techniques were used to characterize the condition of the fins in an effort to identify degradation mechanisms, and the relevant findings are presented in this paper.

  8. 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

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

    International Nuclear Information System (INIS)

    Burdakov, N.S.

    1996-01-01

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

  10. Modeling of beam-target interaction during pulsed electron beam ablation of graphite: Case of melting

    Energy Technology Data Exchange (ETDEWEB)

    Ali, Muddassir, E-mail: mx1_ali@laurentian.ca; Henda, Redhouane

    2017-02-28

    Highlights: • Modeling of ablation stage induced during pulsed electron beam ablation (PEBA). • Thermal model to describe heating, melting and vaporization of a graphite target. • Model results show good accordance with reported data in the literature. - Abstract: A one-dimensional thermal model based on a two-stage heat conduction equation is employed to investigate the ablation of graphite target during nanosecond pulsed electron beam ablation. This comprehensive model accounts for the complex physical phenomena comprised of target heating, melting and vaporization upon irradiation with a polyenergetic electron beam. Melting and vaporization effects induced during ablation are taken into account by introducing moving phase boundaries. Phase transition induced during ablation is considered through the temperature dependent thermodynamic properties of graphite. The effect of electron beam efficiency, power density, and accelerating voltage on ablation is analyzed. For an electron beam operating at an accelerating voltage of 15 kV and efficiency of 0.6, the model findings show that the target surface temperature can reach up to 7500 K at the end of the pulse. The surface begins to melt within 25 ns from the pulse start. For the same process conditions, the estimated ablation depth and ablated mass per unit area are about 0.60 μm and 1.05 μg/mm{sup 2}, respectively. Model results indicate that ablation takes place primarily in the regime of normal vaporization from the surface. The results obtained at an accelerating voltage of 15 kV and efficiency factor of 0.6 are satisfactorily in good accordance with available experimental data in the literature.

  11. Measurements of the thermophysical properties of graphite composites for a neutron target converter

    Energy Technology Data Exchange (ETDEWEB)

    Zhmurikov, E.I. [Budker Institute of Nuclear Physics, 690090 Novosibirsk, SB RAS (Russian Federation); Savchenko, I.V.; Stankus, S.V.; Yatsuk, O.S. [Kutateladze Institute of Thermal Physics, 690090 Novosibirsk, SB RAS (Russian Federation); Tecchio, L.B., E-mail: tecchio@lnl.infn.it [Laboratori Nazionali di Legnaro-Istituto Nazionale di Fisica Nucleare, Legnaro (Italy)

    2012-05-11

    The thermo-physical characteristics of both nuclear, industrial graphite, such as MPG-6 from NIIGRAFIT or SGL composite from SGL Carbon Group, and the first synthesized graphite composite based on a carbon {sup 13}S isotope have been measured from room temperature to 1675 K by laser flash method on an LFA-427 setup from Netzsch (Germany). The results obtained are compared to the previous data of X-Ray analysis and high-resolution electron microscopy and can be used in thermal calculations of the design of a neutron converter graphite target.

  12. Measurements of the thermophysical properties of graphite composites for a neutron target converter

    International Nuclear Information System (INIS)

    Zhmurikov, E.I.; Savchenko, I.V.; Stankus, S.V.; Yatsuk, O.S.; Tecchio, L.B.

    2012-01-01

    The thermo-physical characteristics of both nuclear, industrial graphite, such as MPG-6 from NIIGRAFIT or SGL composite from SGL Carbon Group, and the first synthesized graphite composite based on a carbon 13 S isotope have been measured from room temperature to 1675 K by laser flash method on an LFA-427 setup from Netzsch (Germany). The results obtained are compared to the previous data of X-Ray analysis and high-resolution electron microscopy and can be used in thermal calculations of the design of a neutron converter graphite target.

  13. Targets of operation monitoring

    International Nuclear Information System (INIS)

    Wenger, H.

    1997-01-01

    The condition monitoring system of a nuclear power plant should be an assurance for the user and the authorities that the safety of the installation constantly corresponds to the nominal condition. As soon as deviations from these conditions occur, it should quickly provide fundamentals for corrective measures. It contributes to a high degree of availability of the installation and makes safe and economical operation throughout the whole life of the installation possible. (author) 1 fig

  14. Negative pressure and spallation in graphite targets under nano- and picosecond laser irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Belikov, R S; Khishchenko, K V [Joint Institute for High Temperatures, Russian Academy of Sciences, Moscow (Russian Federation); Krasyuk, I K; Semenov, A Yu; Stuchebryukhov, I A [A M Prokhorov General Physics Institute, Russian Academy of Sciences, Moscow (Russian Federation); Rinecker, T; Schoenlein, A [Goethe University Frankfurt am Main (Germany); Rosmej, O N [GSI Helmholtzzentrum für Schwerionenforschung GmbH, Germany, 64291 Darmstadt, Planckstraße, 1 (Germany); Tomut, M [Technische Universität Darmstadt, Germany, 64289 Darmstadt, Karolinenplatz, 5 (Germany)

    2015-05-31

    We present the results of experiments on the spallation phenomena in graphite targets under shock-wave nano- and picosecond irradiation, which have been performed on Kamerton-T (GPI, Moscow, Russia) and PHELIX (GSI, Darmstadt, Germany) laser facilities. In the range of the strain rates of 10{sup 6} – 10{sup 7} s{sup -1}, the data on the dynamic mechanical strength of the material at rapure (spallation) have been for the first time obtained. With a maximal strain rate of 1.4 × 10{sup 7} s{sup -1}, the spall strength of 2.1 GPa is obtained, which constitutes 64% of the theoretical ultimate tensile strength of graphite. The effect of spallation is observed not only on the rear side of the target, but also on its irradiated (front) surface. With the use of optical and scanning electron microscopes, the morphology of the front and rear surfaces of the targets is studied. By means of Raman scattering of light, the graphite structure both on the target front surface under laser exposure and on its rear side in the spall zone is investigated. A comparison of the dynamic strength of graphite and synthetic diamond is performed. (extreme light fields and their applications)

  15. Time resolved optical emission spectroscopy of cross-beam pulsed laser ablation on graphite targets

    International Nuclear Information System (INIS)

    Sangines, R.; Sanchez Ake, C.; Sobral, H.; Villagran-Muniz, M.

    2007-01-01

    Cross-beam pulsed laser ablation with two delayed lasers is performed on two perpendicular graphite targets. The time delay between lasers is varied by up to 5 μs, and physical changes on the second plasma, due to the interaction with the first generated one, are determined by time resolved optical emission spectroscopy

  16. Development, installation, and initial operation of DIII-D graphite armor tiles

    International Nuclear Information System (INIS)

    Anderson, P.M.; Baxi, C.B.; Reis, E.E.; Smith, J.P.; Smith, P.D.

    1988-04-01

    An upgrade of the DIII-D vacuum vessel protection system has been completed. The ceiling, floor, and inner wall have been armored to enable operation of CIT-relevant doublenull diverted plasmas and to enable the use of the inner wall as a limiting surface. The all- graphite tiles replace the earlier partial coverage armor configuration which consisted of a combination of Inconel tiles and graphite brazed to Inconel tiles. A new all-graphite design concept was chosen for cost and reliability reasons. The 10 minute duration between plasma discharges required the tiles to be cooled by conduction to the water-cooled vessel wall. Using two and three- dimensional analyses, the tile design was optimized to minimize thermal stresses with uniform thermal loading on the plasma-facing surface. Minimizing the stresses around the tile hold-down feature and eliminating stress concentrators were emphasized in the design. The design of the tile fastener system resulted in sufficient hold-down forces for good thermal conductance to the vessel and for securing the tile against eddy current forces. The tiles are made of graphite, and a program to select a suitable grade of graphite was undertaken. Initially, graphites were compared based on published technical data. Graphite samples were then tested for thermal shock capacity in an electron beam test facility at the Sandia National Laboratory (SNLA) in Albuquerque, New Mexico, USA. 4 refs., 6 figs

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

    CERN Document Server

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

    2003-01-01

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

  18. Quality of Graphite Target for Biological/Biomedical/Environmental Applications of 14C-Accelerator Mass Spectrometry

    Science.gov (United States)

    2010-01-01

    Catalytic graphitization for 14C-accelerator mass spectrometry (14C-AMS) produced various forms of elemental carbon. Our high-throughput Zn reduction method (C/Fe = 1:5, 500 °C, 3 h) produced the AMS target of graphite-coated iron powder (GCIP), a mix of nongraphitic carbon and Fe3C. Crystallinity of the AMS targets of GCIP (nongraphitic carbon) was increased to turbostratic carbon by raising the C/Fe ratio from 1:5 to 1:1 and the graphitization temperature from 500 to 585 °C. The AMS target of GCIP containing turbostratic carbon had a large isotopic fractionation and a low AMS ion current. The AMS target of GCIP containing turbostratic carbon also yielded less accurate/precise 14C-AMS measurements because of the lower graphitization yield and lower thermal conductivity that were caused by the higher C/Fe ratio of 1:1. On the other hand, the AMS target of GCIP containing nongraphitic carbon had higher graphitization yield and better thermal conductivity over the AMS target of GCIP containing turbostratic carbon due to optimal surface area provided by the iron powder. Finally, graphitization yield and thermal conductivity were stronger determinants (over graphite crystallinity) for accurate/precise/high-throughput biological, biomedical, and environmental14C-AMS applications such as absorption, distribution, metabolism, elimination (ADME), and physiologically based pharmacokinetics (PBPK) of nutrients, drugs, phytochemicals, and environmental chemicals. PMID:20163100

  19. In situ targeted MRI detection of Helicobacter pylori with stable magnetic graphitic nanocapsules

    Science.gov (United States)

    Li, Yunjie; Hu, Xiaoxiao; Ding, Ding; Zou, Yuxiu; Xu, Yiting; Wang, Xuewei; Zhang, Yin; Chen, Long; Chen, Zhuo; Tan, Weihong

    2017-06-01

    Helicobacter pylori infection is implicated in the aetiology of many diseases. Despite numerous studies, a painless, fast and direct method for the in situ detection of H. pylori remains a challenge, mainly due to the strong acidic/enzymatic environment of the gastric mucosa. Herein, we report the use of stable magnetic graphitic nanocapsules (MGNs), for in situ targeted magnetic resonance imaging (MRI) detection of H. pylori. Several layers of graphene as the shell effectively protect the magnetic core from corrosion while retaining the superior contrast effect for MRI in the gastric environment. Boronic-polyethylene glycol molecules were synthesized and modified on the MGN surface for targeted MRI detection. In a mouse model of H. pylori-induced infection, H. pylori was specifically detected through both T2-weighted MR imaging and Raman gastric mucosa imaging using functionalized MGNs. These results indicated that enhancement of MRI using MGNs may be a promising diagnostic and bioimaging platform for very harsh conditions.

  20. Spatial distribution of carbon species in laser ablation of graphite target

    International Nuclear Information System (INIS)

    Ikegami, T.; Ishibashi, S.; Yamagata, Y.; Ebihara, K.; Thareja, R.K.; Narayan, J.

    2001-01-01

    We report on the temporal evolution and spatial distribution of C 2 and C 3 molecules produced by KrF laser ablation of a graphite target using laser induced fluorescence imaging and optical emission spectroscopy. Spatial density profiles of C 2 were measured using two-dimensional fluorescence in various pressures of different ambient (vacuum, nitrogen, oxygen, hydrogen, helium, and argon) gases at various ablation laser fluences and ablation area. A large yield of C 2 is observed in the central part of the plume and near the target surface and its density and distribution was affected by the laser fluence and ambient gas. Fluorescent C 3 was studied in Ar gas and the yield of C 3 is enhanced at higher gas pressure and longer delay times after ablation

  1. 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)

  2. Laser ablation of a silicon target in chloroform: formation of multilayer graphite nanostructures

    Science.gov (United States)

    Abderrafi, Kamal; García-Calzada, Raúl; Sanchez-Royo, Juan F.; Chirvony, Vladimir S.; Agouram, Saïd; Abargues, Rafael; Ibáñez, Rafael; Martínez-Pastor, Juan P.

    2013-04-01

    With the use of high-resolution transmission electron microscopy, selected area electron diffraction and x-ray photoelectron spectroscopy methods of analysis we show that the laser ablation of a Si target in chloroform (CHCl3) by nanosecond UV pulses (40 ns, 355 nm) results in the formation of about 50-80 nm core-shell nanoparticles with a polycrystalline core composed of small (5-10 nm) Si and SiC mono-crystallites, the core being coated by several layers of carbon with the structure of graphite (the shell). In addition, free carbon multilayer nanostructures (carbon nano-onions) are also found in the suspension. On the basis of a comparison with similar laser ablation experiments implemented in carbon tetrachloride (CCl4), where only bare (uncoated) Si nanoparticles are produced, we suggest that a chemical (solvent decomposition giving rise to highly reactive CH-containing radicals) rather than a physical (solvent atomization followed by carbon nanostructure formation) mechanism is responsible for the formation of graphitic shells. The silicon carbonization process found for the case of laser ablation in chloroform may be promising for silicon surface protection and functionalization.

  3. Computer simulation of sputtering of graphite target in magnetron sputtering device with two zones of erosion

    Directory of Open Access Journals (Sweden)

    Bogdanov R.V.

    2015-03-01

    Full Text Available A computer simulation program for discharge in a magnetron sputtering device with two erosion zones was developed. Basic laws of the graphite target sputtering process and transport of sputtered material to the substrate were taken into account in the Monte Carlo code. The results of computer simulation for radial distributions of density and energy flux of carbon atoms on the substrate (at different values of discharge current and pressure of the working gas confirmed the possibility of obtaining qualitative homogeneous films using this magnetron sputtering device. Also the discharge modes were determined for this magnetron sputtering device, in which it was possible to obtain such energy and density of carbon atoms fluxes, which were suitable for deposition of carbon films containing carbon nanotubes and other nanoparticles.

  4. High temperature vacuum furnace for the preparation of graphite targets for 14C dating by tandem accelerator mass spectrometry

    International Nuclear Information System (INIS)

    Lowe, D.C.; Bristow, P.; Judd, W.J.

    1985-02-01

    A simple and reliable furnace design capable of producing temperatures of up to 2800 deg. C is presented. The furnace has been specifically designed for the rapid and reliable production of graphite targets for 14 C dating purposes but may be used in a variety of applications requiring high temperatures under vacuum conditions

  5. Spherical cauliflower-like carbon dust formed by interaction between deuterium plasma and graphite target and its internal structure

    Energy Technology Data Exchange (ETDEWEB)

    Ohno, N. [Department of Energy Engineering and Science, Graduate School of Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan)], E-mail: ohno@ees.nagoya-u.ac.jp; Yoshimi, M. [Department of Energy Engineering and Science, Graduate School of Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan); Tokitani, M. [National Institute for Fusion Science, Oroshi 322-6, Toki 509-5292 (Japan); Takamura, S. [Department of Electronics, Aichi Institute of Technology, Yakusa-cho, Toyota 470-0392 (Japan); Tokunaga, K.; Yoshida, N. [Research Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan)

    2009-06-15

    Simulated experiments to produce carbon dust particles with cauliflower structure have been performed in a liner plasma device, NAGDIS-II by exposing high density deuterium plasma to a graphite sample (IG-430U). Formation of carbon dust depends on the surface temperature and the incident ion energy. At a surface temperature 600-700 K, a lot of isolated spherical dust particles are observed on the graphite target. The internal structure of an isolated dust particle was observed with Focused Ion Beam (FIB) system and Transmission Electron Microscope (TEM) in detail. FIB analysis clearly shows there exist honey-combed cell structure with thin carbon walls in the dust particle and the dust particle grows from the graphite surface. TEM image also shows that the dust particle is made of amorphous carbon with crystallized grains with diameters of 10-50 nm.

  6. Fragmentation of 400 MeV/u {sup 12}C on a thin graphite target

    Energy Technology Data Exchange (ETDEWEB)

    Schuy, Christoph

    2015-07-01

    Detailed understanding of high energetic heavy ions interacting with matter is of great interest in basic research and applied physics especially in radiotherapy and space radioprotection. Radiotherapy with carbon ions showed great success especially in the treatment of deep seated tumors due to the favorable depth-dose profile and increased biological effectiveness compared to photons or protons. Due to nuclear interactions between the primary beam and the patient's body, usually only 50% of the carbon ions will reach the target location. Thus, a detailed knowledge of the changes in the radiation field is required for delivering a successful treatment. The radiation environment in space is composed of high energy charged particles and can lead to serious health risks for astronauts. The assessment and mitigation of radioinduced health complications cannot be accomplished without a good understanding of the interaction of the mixed radiation field with e.g. the hull of the spaceship or lunar soil. In this work the fragmentation of 400 MeV/u {sup 12}C on a thin graphite target was investigated. The resulting angular yield distributions and differential energy spectra of charged and uncharged particles are presented and compared to two different Monte Carlo codes (PHITS and GEANT4).

  7. Nuclear graphite development, operational problems, and resolution of these problems at the Hanford production reactors

    International Nuclear Information System (INIS)

    Morgan, W.C.

    1996-01-01

    This paper chronicles the history of the Hanford Production Reactor, from the initial design considerations for B, D, and F Reactors through the selection of the agreed method for safe disposal of the decommissioned reactors. The operational problems that challenged the operations and support staff of each new generation of production reactors, the engineering actions an operational changes that alleviated or resolved the immediate problems, the changes in reactor design and design-bases for the next generation of production reactors, and the changes in manufacturing variables that resulted in new ''improved'' grades of nuclear graphites for use in the moderators of the Hanford Production Reactors are reviewed in the context of the existing knowledge-base and the mission-driven priorities on the time. 14 refs, 6 figs, 3 tabs

  8. ISAC target operation with high proton currents

    CERN Document Server

    Dombsky, M; Schmor, P; Lane, M

    2003-01-01

    The TRIUMF-ISAC facility target stations were designed for ISOL target irradiations with up to 100 mu A proton beam currents. Since beginning operation in 1998, ISAC irradiation currents have progressively increased from initial values of approx 1 mu A to present levels of up to 40 mu A on refractory metal foil targets. In addition, refractory carbide targets have operated at currents of up to 15 mu A for extended periods. The 1-40 mu A operational regime is achieved by tailoring each target to the thermal requirements dictated by material properties such as beam power deposition, thermal conductivity and maximum operating temperature of the target material. The number of heat shields on each target can be varied in order to match the effective emissivity of the target surface for the required radiative power dissipation. Targets of different thickness, surface area and volume have been investigated to study the effect of diffusion and effusion delays on the yield of radioisotopes. For yields of short-lived p...

  9. Polarized proton target-IV. Operations manual

    International Nuclear Information System (INIS)

    Hill, D.; Fletcher, O.; Moretti, A.; Onesto, F.

    1976-01-01

    Standard operating procedures are presented for the vacuum, cryogenic, and electronic systems of a polarized proton target. The systems are comprised of (1) a target cryostat; (2) a 4 He pumping system; (3) a 3 He pumping system; (4) a microwave system; (5) a magnet and power supply; (6) a computerized polarization monitor; and (7) miscellaneous auxiliary equipment

  10. Confinement effects of shock waves on laser-induced plasma from a graphite target

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Feiling; Liang, Peipei; Yang, Xu; Cai, Hua; Wu, Jiada; Xu, Ning; Ying, Zhifeng; Sun, Jian, E-mail: jsun@fudan.edu.cn [Shanghai Ultra-Precision Optical Manufacturing Engineering Center, Department of Optical Science and Engineering, Fudan University, Shanghai 200433 (China)

    2015-06-15

    The spatial confinement effects of shock waves on the laser-induced plasma (LIP) from a graphite target in air were studied by probe beam deflection (PBD) measurements and optical emission spectroscopy (OES). A clear relationship between the confinement of the LIP by the shock wave and the effects on the LIP emission was observed, and the underlying mechanisms are discussed. PBD monitoring revealed that the laser-ablation induced shock wave could be well analogized to the shock wave generated by a point explosion and would be reflected by a block. OES measurements indicated that the optical emission of the LIP exhibited significant variations with the block placement. A first enhancement and then a fast decay of CN molecular emission as well as a suppression of carbon atomic emission were observed in the presence of the block. The results revealed that the reflected shock wave spatially confined the expansion of the LIP and compressed the LIP after encountering it, pushing back the species of the LIP and changing the density of the LIP species including luminous carbon atoms and CN molecules. It is suggested that the change of the LIP emission is attributed to the density variation of the LIP species due to the compression of the LIP and the reactions occurring in the plasma.

  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. Bomb conversion of CO2 to graphite for radiocarbon measurement by tandem accelerator mass spectrometry (TAMS)

    International Nuclear Information System (INIS)

    Lowe, D.C.; Judd, W.J.

    1985-01-01

    TAMS has extended the scope of radiocarbon dating but improved sample target preparation techniques are still required. In this paper a quick and reliable method for the conversion of CO 2 into graphite targets suitable for radiocarbon dating is presented. In the outlined procedure a single operator can produce 20 graphite targets a day. (author)

  13. Computer programs in BASIC language for graphite furnace atomic absorption using the method of additions. Part 1. Operating instructions

    International Nuclear Information System (INIS)

    Boyle, W.G. Jr.; Ryan, D.P.

    1979-01-01

    These instructions describe how to use BASIC language programs to process data from atomic absorption spectrophotometers using the graphite furnace and the method of additions calibration technique. The instructions cover loading the programs, responding to computer prompts, choosing among various options for processing the data, performing operations with an automatic sampler, and producing reports. How the programs interact with each other is also explained. Examples of computer/operator dialogue are presented for typical cases. In addition, a concise set of operating instructions is included as an appendix

  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. 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)

  16. 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

  17. Investigations of fast neutron production by 190 GeV/c muon interactions on graphite target

    CERN Document Server

    Chazal, V; Cook, B; Henrikson, H; Jonkmans, G; Paic, A; Mascarenhas, N; Vogel, P; Vuilleumier, J L

    2002-01-01

    The production of fast neutrons (1 MeV - 1 GeV) in high energy muon-nucleus interactions is poorly understood, yet it is fundamental to the understanding of the background in many underground experiments. The aim of the present experiment (CERN NA55) was to measure spallation neutrons produced by 190 GeV/c muons scattering on carbon target. We have investigated the energy spectrum and angular distribution of spallation neutrons, and we report the result of our measurement of the neutron production differential cross section.

  18. Simulations of the Full Impact of the LHC Beam on Solid Copper and Graphite Targets

    CERN Document Server

    Tahir, Naeem; Lomonosov, Igor; Shutov, Alexander; Piriz, Roberto; Schmidt, Ruediger

    2010-01-01

    The CERN Large Hadron Collider (LHC) is by far the most powerful accelerator in the world. It is a 26.8 km circumference proton synchrotronwith 1232 superconducting magnets, accelerating two counter–rotating proton beams. When this accelerator will achieve its full capacity, each beam will consist of a bunch train with 2808 bunches and each bunch comprising of 1.15 × 1011 7 TeV protons. The bunch length will be 0.5 ns and two neighboring bunches will be separated by 25 ns while intensity distribution in the radial direction will be Gaussian with a standard deviation, σ = 0.2 mm. In the center of the physics detectors the beam will be focused to a much smaller size, down to a σ of 20 μm. The total duration of the beam will be of the order of 89 μs and the total number of protons in the beam will be 3 × 1014 which is equivalent to 362 MJ energy, sufficient to melt 500 kg copper. Safety of operation is a very important issue when working with such extremely powerful beams. An accidental loss of even a sm...

  19. Numerical simulations of energy deposition caused by 50 MeV—50 TeV proton beams in copper and graphite targets

    Science.gov (United States)

    Nie, Y.; Schmidt, R.; Chetvertkova, V.; Rosell-Tarragó, G.; Burkart, F.; Wollmann, D.

    2017-08-01

    The conceptual design of the Future Circular Collider (FCC) is being carried out actively in an international collaboration hosted by CERN, for the post-Large Hadron Collider (LHC) era. The target center-of-mass energy of proton-proton collisions for the FCC is 100 TeV, nearly an order of magnitude higher than for LHC. The existing CERN accelerators will be used to prepare the beams for FCC. Concerning beam-related machine protection of the whole accelerator chain, it is critical to assess the consequences of beam impact on various accelerator components in the cases of controlled and uncontrolled beam losses. In this paper, we study the energy deposition of protons in solid copper and graphite targets, since the two materials are widely used in magnets, beam screens, collimators, and beam absorbers. Nominal injection and extraction energies in the hadron accelerator complex at CERN were selected in the range of 50 MeV-50 TeV. Three beam sizes were studied for each energy, corresponding to typical values of the betatron function. Specifically for thin targets, comparisons between fluka simulations and analytical Bethe equation calculations were carried out, which showed that the damage potential of a few-millimeter-thick graphite target and submillimeter-thick copper foil can be well estimated directly by the Bethe equation. The paper provides a valuable reference for the quick evaluation of potential damage to accelerator elements over a large range of beam parameters when beam loss occurs.

  20. Determination of fluorine in copper concentrate via high-resolution graphite furnace molecular absorption spectrometry and direct solid sample analysis - Comparison of three target molecules.

    Science.gov (United States)

    Cadorim, Heloisa R; de Gois, Jefferson S; Borges, Aline R; Vale, Maria Goreti R; Welz, Bernhard; Gleisner, Heike; Ott, Christina

    2018-01-01

    The chemical composition of complex inorganic materials, such as copper concentrate, may influence the economics of their further processing because most smelters, and particularly the producers of high-purity electrolyte copper, have strict limitations for the permissible concentration of impurities. These components might be harmful to the quality of the products, impair the production process and be hazardous to the environment. The goal of the present work is the development of a method for the determination of fluorine in copper concentrate using high-resolution graphite furnace molecular absorption spectrometry and direct solid sample analysis. The molecular absorption of the diatomic molecule CaF was measured at 606.440nm. The molecule CaF was generated by the addition of 200µg Ca as the molecule-forming reagent; the optimized pyrolysis and vaporization temperatures were 900°C and 2400°C, respectively. The characteristic mass and limit of detection were 0.5ng and 3ng, respectively. Calibration curves were established using aqueous standard solutions containing the major components Cu, Fe, S and the minor component Ag in optimized concentrations. The accuracy of the method was verified using certified reference materials. Fourteen copper concentrate samples from Chile and Australia were analyzed to confirm the applicability of the method to real samples; the concentration of fluorine ranged from 34 to 5676mgkg -1 . The samples were also analyzed independently at Analytik Jena by different operators, using the same equipment, but different target molecules, InF and GaF, and different operating conditions; but with a few exceptions, the results agreed quite well. The results obtained at Analytik Jena using the GaF molecule and our results obtained with CaF, with one exception, were also in agreement with the values informed by the supplier of the samples, which were obtained using ion selective electrode potentiometry after alkaline fusion. A comparison will

  1. Scale-up of high specific activity {sup 186g}Re production using graphite-encased thick {sup 186}W targets and demonstration of an efficient target recycling process

    Energy Technology Data Exchange (ETDEWEB)

    Balkin, Ethan R.; Gagnon, Katherine; Dorman, Eric [Washington Univ., Seattle, WA (United States). Dept. of Radiation Oncology; and others

    2017-07-01

    Production of high specific activity {sup 186g}Re is of interest for development of theranostic radiopharmaceuticals. Previous studies have shown that high specific activity {sup 186g}Re can be obtained by cyclotron irradiation of enriched {sup 186}W via the {sup 186}W(d,2n){sup 186g}Re reaction, but most irradiations were conducted at low beam currents and for short durations. In this investigation, enriched {sup 186}W metal targets were irradiated at high incident deuteron beam currents to demonstrate production rates and contaminants produced when using thick targets. Full-stopping thick targets, as determined using SRIM, were prepared by uniaxial pressing of powdered natural abundance W metal or 96.86% enriched {sup 186}W metal encased between two layers of graphite flakes for target material stabilization. An assessment of structural integrity was made on each target preparation. To assess the performance of graphite-encased thick {sup 186}W metal targets, along with the impact of encasing on the separation chemistry, targets were first irradiated using a 22 MeV deuteron beam for 10 min at 10, 20, and 27 μA, with an estimated nominal deuteron energy of 18.7 MeV on the {sup 186}W target material (after energy degradation correction from top graphite layer). Gamma-ray spectrometry was performed post EOB on all targets to assess production yields and radionuclidic byproducts. The investigation also evaluated a method to recover and recycle enriched target material from a column isolation procedure. Material composition analyses of target materials, pass-through/wash solutions and recycling process isolates were conducted with SEM, FTIR, XRD, EDS and ICP-MS spectrometry. To demonstrate scaled-up production, a graphite-encased {sup 186}W target made from recycled {sup 186}W was irradiated for ∝2 h with 18.7 MeV deuterons at a beam current of 27 μA to provide 0.90 GBq (24.3 mCi) of {sup 186g}Re, decay-corrected to the end of bombardment. ICP-MS analysis of the

  2. Fabrication and characterization of tungsten and graphite based PFC for divertor target elements of ITER like tokamak application

    Energy Technology Data Exchange (ETDEWEB)

    Khirwadkar, S.S., E-mail: sameer@ipr.res.in [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Singh, K.P.; Patil, Y.; Khan, M.S.; Buch, J.J.U.; Patel, Alpesh; Tripathi, Sudhir [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Jaman, P.M.; Rangaraj, L.; Divakar, C. [Materials Science Division, National Aerospace Laboratories, CSIR, Bangalore, Karnataka (India)

    2011-10-15

    The development of the fabrication technology of macro-brush configuration of tungsten (W) and carbon (graphite and CFC) plasma facing components (PFCs) for ITER like tokamak application is presented. The fabrication of qualified joint of PFC is a requirement for fusion tokamak. Vacuum brazing method has been employed for joining of W/CuCrZr and C/CuCrZr. Oxygen free high conductivity (OFHC) copper casting on W tiles was performed followed by machining, polishing and ultrasonic cleaning of the samples prior to vacuum brazing. The W/CuCrZr and graphite/CuCrZr based test mockups were vacuum brazed using silver free alloys. The mechanical shear and tensile strengths were evaluated for the W/CuCrZr and graphite/CuCrZr brazed joint samples. The micro-structural examination of the joints showed smooth interface. The details of fabrication and characterization procedure for macro-brush tungsten and carbon based PFC test mockups are presented.

  3. Multiple operating system rotation environment moving target defense

    Science.gov (United States)

    Evans, Nathaniel; Thompson, Michael

    2016-03-22

    Systems and methods for providing a multiple operating system rotation environment ("MORE") moving target defense ("MTD") computing system are described. The MORE-MTD system provides enhanced computer system security through a rotation of multiple operating systems. The MORE-MTD system increases attacker uncertainty, increases the cost of attacking the system, reduces the likelihood of an attacker locating a vulnerability, and reduces the exposure time of any located vulnerability. The MORE-MTD environment is effectuated by rotation of the operating systems at a given interval. The rotating operating systems create a consistently changing attack surface for remote attackers.

  4. Polarized proton Target-III operators manual, revision A

    International Nuclear Information System (INIS)

    Hill, D.; Moretti, A.; Onesto, F.; Rynes, P.

    1976-04-01

    A revision is given of a manual containing standard operating procedures for the vacuum, cryogenic, and electronic systems of a polarized proton target. The discussion includes the target cryostat, the 3 He and 4 He pumping systems, remote monitors and controls, the microwave system, the magnet and power supply, the computerized polarization monitor, the 4 He liquifier and gas recovery system, and miscellaneous auxiliary equipment

  5. 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

  6. Numerical simulations of energy deposition caused by 50 MeV—50 TeV proton beams in copper and graphite targets

    Directory of Open Access Journals (Sweden)

    Y. Nie

    2017-08-01

    Full Text Available The conceptual design of the Future Circular Collider (FCC is being carried out actively in an international collaboration hosted by CERN, for the post–Large Hadron Collider (LHC era. The target center-of-mass energy of proton-proton collisions for the FCC is 100 TeV, nearly an order of magnitude higher than for LHC. The existing CERN accelerators will be used to prepare the beams for FCC. Concerning beam-related machine protection of the whole accelerator chain, it is critical to assess the consequences of beam impact on various accelerator components in the cases of controlled and uncontrolled beam losses. In this paper, we study the energy deposition of protons in solid copper and graphite targets, since the two materials are widely used in magnets, beam screens, collimators, and beam absorbers. Nominal injection and extraction energies in the hadron accelerator complex at CERN were selected in the range of 50 MeV–50 TeV. Three beam sizes were studied for each energy, corresponding to typical values of the betatron function. Specifically for thin targets, comparisons between fluka simulations and analytical Bethe equation calculations were carried out, which showed that the damage potential of a few-millimeter-thick graphite target and submillimeter-thick copper foil can be well estimated directly by the Bethe equation. The paper provides a valuable reference for the quick evaluation of potential damage to accelerator elements over a large range of beam parameters when beam loss occurs.

  7. Effects-Based Targeting: Application in Operation Desert Storm and Operation Iraqi Freedom

    National Research Council Canada - National Science Library

    Hansbarger, Thomas

    2004-01-01

    .... The United States military must be able to apply effects-based targeting to capitalize on improved capabilities in operational fires and application of national resources against a dynamic, adapting enemy...

  8. 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

  9. Polarized proton target-III. Operations manual, revision B

    International Nuclear Information System (INIS)

    Hill, D.; Moretti, A.; Onesto, F.; Rynes, P.

    1978-01-01

    The manual presented contains certain standard operating procedures for the vacuum, cryogenic, and electronic systems of PPT-III. In total, these systems comprise the following major divisions: (1) the target cryostat; (2) the 4 He pumping system; (3) the 3 He pumping system; (4) the remote monitors and controls; (5) the microwave system; (6) the magnet and power supply; (7) the computerized polarization monitor; (8) the 4 He liquefier and gas recovery system; and (9) miscellaneous auxiliary equipment

  10. The Effect of Thermal Annealing on the Optical Properties of a-SiC:H Films Produced by DC Sputtering Methods: I. Graphite Target Case.

    Directory of Open Access Journals (Sweden)

    Lusitra Munisa

    2003-04-01

    Full Text Available A study of the annealing effect on optical properties and disorder of hydrogenated amorphous silicon carbon (a-SiC:H films was undertaken. The films were prepared by sputtering technique using graphite target and silicon wafer in argon and hydrogen gas mixture, and then characterized by uv-vis (ultra violet-visible spectroscopy before and after annealing. Index of refraction n and absorption coefficient α of films have been determined from measurements of transmittance. The optical gap show small variation with annealing temperature, increasing with increasing annealing temperature up to 500 °C. An increase of annealing temperature leads to reduced film density and the amorphous network disorder. The experimental results are discussed in terms of deposition condition and compared to other experimental results.

  11. RTNS-II [Rotating Target Neutron Source II] operational summary

    International Nuclear Information System (INIS)

    Heikkinen, D.W.

    1988-09-01

    The Rotating Target Neutron Source II facility (RTNS-II) operated for over nine years. Its purpose was to provide high intensities of 14 MeV neutrons for materials studies in the fusion energy program. For the period from 1982-1987, the facility was supported by both the US (Department of Energy) and Japan (Ministry of Education, Culture, and Science). RTNS-II contains two accelerator-based neutron sources which use the T(d,n) 4 He reaction. In this paper, we will summarize the operational history of RTNS-II. Typical operating parameters are given. In addition, a brief description of the experimental program is presented. The current status and future options for the facility are discussed. 7 refs., 5 tabs

  12. 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

  13. 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.

  14. 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.

  15. 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)

  16. Numerical simulations of energy deposition caused by 50 MeV—50 TeV proton beams in copper and graphite targets

    CERN Document Server

    Nie, Y; Chetvertkova, V; Rosell-Tarrago, G; Burkart, F; Wollmann, D

    2017-01-01

    The conceptual design of the Future Circular Collider (FCC) is being carried out actively in an international collaboration hosted by CERN, for the post–Large Hadron Collider (LHC) era. The target center-of-mass energy of proton-proton collisions for the FCC is 100 TeV, nearly an order of magnitude higher than for LHC. The existing CERN accelerators will be used to prepare the beams for FCC. Concerning beam-related machine protection of the whole accelerator chain, it is critical to assess the consequences of beam impact on various accelerator components in the cases of controlled and uncontrolled beam losses. In this paper, we study the energy deposition of protons in solid copper and graphite targets, since the two materials are widely used in magnets, beam screens, collimators, and beam absorbers. Nominal injection and extraction energies in the hadron accelerator complex at CERN were selected in the range of 50 MeV–50 TeV. Three beam sizes were studied for each energy, corresponding to typical values ...

  17. Preliminary design implications of SSC fixed-target operation

    International Nuclear Information System (INIS)

    Zisman, M.S.

    1984-06-01

    This paper covers some of the accelerator physics issues relevant to a possible fixed-target operating mode for the Superconducting Super Collider (SSC). In the brief time available, no attempt has been made to design this capability into the SSC. Rather, I have tried to evaluate what the performance of such a machine might be, and to indicate the hardware implications and extraction considerations that would be part of an actual design study. Where appropriate, parameters and properties of the present LBL design for the SSC have been used; these should be taken as being representative of the general class of small-aperture, high-field colliders being considered by the accelerator physics community. Thus, the numerical examples given here must ultimately be reexamined in light of the actual parameters of the particular accelerator being considered

  18. 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.

  19. 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.

  20. Clinical target volume delineation in glioblastomas: pre-operative versus post-operative/pre-radiotherapy MRI

    Science.gov (United States)

    Farace, P; Giri, M G; Meliadò, G; Amelio, D; Widesott, L; Ricciardi, G K; Dall'Oglio, S; Rizzotti, A; Sbarbati, A; Beltramello, A; Maluta, S; Amichetti, M

    2011-01-01

    Objectives Delineation of clinical target volume (CTV) is still controversial in glioblastomas. In order to assess the differences in volume and shape of the radiotherapy target, the use of pre-operative vs post-operative/pre-radiotherapy T1 and T2 weighted MRI was compared. Methods 4 CTVs were delineated in 24 patients pre-operatively and post-operatively using T1 contrast-enhanced (T1PRECTV and T1POSTCTV) and T2 weighted images (T2PRECTV and T2POSTCTV). Pre-operative MRI examinations were performed the day before surgery, whereas post-operative examinations were acquired 1 month after surgery and before chemoradiation. A concordance index (CI) was defined as the ratio between the overlapping and composite volumes. Results The volumes of T1PRECTV and T1POSTCTV were not statistically different (248 ± 88 vs 254 ± 101), although volume differences >100 cm3 were observed in 6 out of 24 patients. A marked increase due to tumour progression was shown in three patients. Three patients showed a decrease because of a reduced mass effect. A significant reduction occurred between pre-operative and post-operative T2 volumes (139 ± 68 vs 78 ± 59). Lack of concordance was observed between T1PRECTV and T1POSTCTV (CI = 0.67 ± 0.09), T2PRECTV and T2POSTCTV (CI = 0.39 ± 0.20) and comparing the portion of the T1PRECTV and T1POSTCTV not covered by that defined on T2PRECTV images (CI = 0.45 ± 0.16 and 0.44 ± 0.17, respectively). Conclusion Using T2 MRI, huge variations can be observed in peritumoural oedema, which are probably due to steroid treatment. Using T1 MRI, brain shifts after surgery and possible progressive enhancing lesions produce substantial differences in CTVs. Our data support the use of post-operative/pre-radiotherapy T1 weighted MRI for planning purposes. PMID:21045069

  1. Proton irradiated graphite grades for a long baseline neutrino facility experiment

    International Nuclear Information System (INIS)

    Simos, N.; Nocera, P.; Zwaska, R.; Mokhov, N.

    2017-01-01

    In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF) of the Deep Underground Neutrino Experiment (DUNE) four graphite grades were irradiated with protons in the energy range of 140–180 MeV, to peak fluence of ~6.1×10"2"0 p/cm"2 and irradiation temperatures between 120–200 °C. The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a) comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b) understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young’s modulus. The proton fluence level of ~10"2"0 cm"-"2 where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite revealed for the first time the similarity in

  2. Proton irradiated graphite grades for a long baseline neutrino facility experiment

    Science.gov (United States)

    Simos, N.; Nocera, P.; Zhong, Z.; Zwaska, R.; Mokhov, N.; Misek, J.; Ammigan, K.; Hurh, P.; Kotsina, Z.

    2017-07-01

    In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF) of the Deep Underground Neutrino Experiment (DUNE) four graphite grades were irradiated with protons in the energy range of 140-180 MeV, to peak fluence of ˜6.1 ×1020 p /cm2 and irradiation temperatures between 120 - 200 °C . The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a) comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b) understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young's modulus. The proton fluence level of ˜1020 cm-2 where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite revealed for the first time the similarity in

  3. 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

  4. 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

  5. [Role-specific targets and teamwork in the operating room].

    Science.gov (United States)

    Hoeper, K; Kriependorf, M; Felix, C; Nyhuis, P; Tecklenburg, A

    2017-12-01

    The primary goal of a surgical team is the successful performance of an operation on a patien; however, this primary goal can show discrepancies from the goals of individual team members. The main causes for differences of interests can be variations in subjective preferences and organizational differences. Subjective preferences are due to the values held by those involved. These values are of an intrinsic nature and therefore difficult to change. Another reason for individual goals is that hospitals and universities are professional bureaucracies. Experts working in professional bureaucracies are known to identify themselves to a greater extent with their respective profession than with their institution; however, teams in the operating room (OR) have to work together in multidisciplinary teams. The main goal of this analysis is to document role-specific targets and motivations within teams. This was a case study at a university hospital with 40 operating rooms. The data collection resulted from the three pillars of the goal documentation instrument, which includes expert interviews, a utility analysis and card placement as a basis for communicative validation. The results were analyzed with a systematic method as a qualitative content analysis. The four-pillar success model, which maps aspects of a successful hospital, was used as a deductive coding scheme. The four pillars represent the level of medical quality (process, structure and outcome quality), economy and efficiency, client satisfaction (patients and referring physicians) and employee satisfaction. At a university hospital an additional focus is on research and teaching. In addition to the four pillar success model as a deductive coding scheme, an inductive coding scheme was introduced. Approximately 10% of the employees from each professional group (surgeons, anesthesiologists, OR nurses, nurse anesthetists) were interviewed resulting in 65 interviews overall. The interviews were conducted

  6. 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.

  7. 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

  8. Temperature distribution in graphite during annealing in air cooled reactors

    International Nuclear Information System (INIS)

    Oliveira Avila, C.R. de.

    1989-01-01

    A model for the evaluation temperature distributions in graphite during annealing operation in graphite. Moderated an-cooled reactors, is presented. One single channel and one dimension for air and graphite were considered. A numerical method based on finite control volumes was used for partioning the mathematical equations. The problem solution involves the use of unsteady equations of mass, momentum and energy conservation for air, and energy conservation for graphite. The source term was considered as stored energy release during annealing for describing energy conservation in the graphite. The coupling of energy conservation equations in air and graphite is performed by the heat transfer term betwen air and graphite. The results agree with experimental data. A sensitivity analysis shown that the termal conductivity of graphite and the maximum inlet channel temperature have great effect on the maximum temperature reached in graphite during the annealing. (author)

  9. Time Sensitive Targeting: Overcoming the Intelligence Gap in Interagency Operations

    National Research Council Canada - National Science Library

    Hewitt, Mark

    2003-01-01

    The Central Intelligence Agency's attack on a group of terrorists in Yemen epitomized the agency's short-notice capability to detect, track, and destroy a highly mobile and fleeting target of opportunity. The U.S...

  10. 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

  11. 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)

  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. 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)

  14. 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

  15. 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

  16. Targeted Killing: Managing American Perceptions On Unmanned Aerial Vehicle Operations

    Science.gov (United States)

    2016-02-01

    Requirements Proposal Advisor: Dr. Patricia Williams Lessane Project Advisor: Dr. Andrew Niesiobedzki Maxwell AFB, AL February 2016...epistemology of remote fighting." Ethics and Information Technology 15. no. 2. 87-98. Cullen , Peter. 2008. "The Role of Targeted Killing in the...in the Sky." New Statesman 19-25. June. 48. Patterson, Margot. 2015. "Are We Safer." America 212. no. 204. 12. Raven-Hansen, William C. Banks and

  17. Development of the HERA-Β-target-control system and study of target operation at the HERA storage ring

    International Nuclear Information System (INIS)

    Issever, S.

    2001-03-01

    The HERA-B experiment investigates the physics of heavy quarks, which are produced in pN reactions of the 920 GeV protons of HERA with the HERA-B internal fixed target. It consists out of eight wires, which surround the proton beam from four sides and is a high luminosity particle source. As being the closest mechanical device to the proton beam, it has to be operated very carefully and thus needs a secure and automatic control system, which additionally must be efficient and reliable to guarantee an efficient HERA-B data taking. The implementation of the target control system and its performance as well as dedicated studies of target-beam physics are presented. These include the measurement of the aperture limitation, usual target operational position, target efficiency, target independent proton loss in HERA and the scrape velocity of the target. The source of rate fluctuations is investigated in detail; among many dependencies environmental noise has a major impact on the rate fluctuations. Further studies include the analysis of beam position fluctuations and its correlation to the rate fluctuations; the rate changes about a factor of 2, if the beam is changing its position by 10 μm. This rate sensitivity is also verified directly by means of step function measurements. Furthermore the step function measurements can be used to study target-beam dynamics on time scales as short as a second. Experiments to reduce the rate sensitivity - the so called beam tail shaping measurements - are presented as well. During target operation a current, which is proportional to the interaction rate, is measured and used to determine the rate of each single wire. It is shown, that the source of this current is delta electron production. Finally the multiwire performance of the target control system is presented. (orig.)

  18. The ISIS operation: Robotics repair work on the CHINON A3 natural uranium, carbon dioxide cooled, graphite moderated reactor

    International Nuclear Information System (INIS)

    Hilmoine, R.M.E.

    1989-01-01

    After describing the upper internal support structures of the CHINON A3 reactor, the problems resulting from their degradation due to corrosion and to the difficulties of the ISIS operation are presented here. The repair method is as follows: all tools and repair parts reach the working area by the feeding-pipes drilled through the 7 m thick concrete vessel surrounding the reactor core; the robots handle into the reactor, the tool heads and the repair parts which are automatically positioned and welded around the corroded structure, thus restoring the support of measurement devices. The parts are either linked together or to the existing structure by means of 2 studs of 12 mm in diameter. The different phases to sort out a problem are: in-core topography, reconforming of the full-scale mock-up with the repair area, learning on this mock-up and in-core repair. The technical specificities of the robots used are the following: they have an 11 meter long, 0.22 meter across telescopic mast with jointed arms reaching a radius of 2.7 m. Then the useful load is 70 daN and the repeatability 0.1 mm. Different tool heads can be handled by the robot: telemeter and laser reconstruction: it allows to locate the in core points and to materialize them on the mock-up by a laser crossed-beams locating technique; scouring: it cleans the corroded parts of the structures before welding; welding: it allows the parts handling and the carried studs welding; screwing; tensile test: carried out when the stud welds are defective. A high level computerized control system is organized around a central unit which calculates the displacements of robots and synchronises the actions of different tools by communicating with several local units. A 100,000 hour designing, a 200,000 hour building and assembling and a 450,000 hour operating on working area were necessary to repair 15 out of the 102 corroded structures by fitting and welding 205 repair parts. 10 figs

  19. 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

  20. Operational Benefits of Meeting California’s Energy Storage Targets

    Energy Technology Data Exchange (ETDEWEB)

    Josh Eichman, Paul Denholm, Jennie Jorgenson, Udi Helman

    2015-12-01

    In October 2013, the California Public Utilities Commission (CPUC) issued rules for its jurisdictional utilities to procure a minimum of 1,325 megawatts (MW) of energy storage systems by 2020. The goal of this study is to examine the operational value of this storage portfolio in California and the rest of the Western Electricity Coordinating Council (WECC) region. Modeled results show that the storage portfolio, when providing energy and operating reserves, reduces the total WECC-wide production costs by $78 million per year in the 33% scenario. This value increases to $144 million per year in the 40% scenario, primarily because of the increase in off-peak and peak price differences that are due to additional solar generation. These values are equivalent to $59/kW-year for the storage portfolio for the 33% scenario and $109/kW-year for the 40% scenario.

  1. 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

  2. Change in properties of graphite on stake of Obninsk NPP

    International Nuclear Information System (INIS)

    Virgul'ev, Yu.S.; Gundorov, V.V.; Kalyagina, I.P.; Belinskaya, N.T.; Dolgov, V.V.; Komissarov, O.V.; Stuzhnev, Yu.A.

    1997-01-01

    The results of testing the graphite from the AM-1 reactor masonry at the Obninsk NPP for its operation period are discussed. It is shown that the masonry graphite state after 42 years of the reactor operation remains satisfactory in the most cells inspected. Separate cells requiring a repair resulted from oxidation are characterized by strength decreased by several times. The laws of radiation changes in graphite properties are analyzed. The conclusion on possibility of the further masonry operation is drawn

  3. 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

  4. Numerical investigations of the WASA pellet target operation and proposal of a new technique for the PANDA pellet target

    Energy Technology Data Exchange (ETDEWEB)

    Varentsov, Victor L., E-mail: v.varentsov@gsi.de [Institute for Theoretical and Experimental Physics, B. Cheremushkinskaya 25, 117218 Moscow (Russian Federation)

    2011-08-01

    The conventional nozzle vibration technique of the hydrogen micro-droplet generation that is supposed to be used for internal pellet target production for the future PANDA experiment at the international FAIR facility in Darmstadtfor is described. The operation of this technique has been investigated by means of detailed computer simulations. Results of calculations for the geometry and operation conditions of the WASA pellet generator are presented and discussed. We have found that for every given pellet size, there is a set of operation parameters where the efficiency of the WASA hydrogen pellet target operation is considerably increased. Moreover, the results of presented computer simulations clearly show that the future PANDA pellet target setup can be realized with the use of much smaller (and cheaper) vacuum pumps than those used at present in the WASA hydrogen pellet target. To qualitatively improve the PANDA hydrogen pellet target performance we have proposed the use of a novel flow focusing method of Ganan-Calvo and Barreto (1997,1999) combined with the use of conventional vacuum injection capillary. Possibilities of this approach for the PANDA pellet target production have been also explored by means of computer simulations. The results of these simulations show that the use of this new approach looks very promising and in particular, there is no need here to use of expensive ultra-pure hydrogen to prevent nozzle clogging or freezing up due to impurities and it will allow simple, fast, smooth and a wide range of change of pellet sizes in accordance with requirements of different experiments at the PANDA detector. In this article we also propose and describe the idea of a new technique to break up a liquid microjet into microdroplets using a process of liquid jet evaporation under pulsed laser beam irradiation. This technique should be experimentally checked before it may be used in the design of the future PANDA pellet target setup.

  5. Numerical investigations of the WASA pellet target operation and proposal of a new technique for the PANDA pellet target

    International Nuclear Information System (INIS)

    Varentsov, Victor L.

    2011-01-01

    The conventional nozzle vibration technique of the hydrogen micro-droplet generation that is supposed to be used for internal pellet target production for the future PANDA experiment at the international FAIR facility in Darmstadtfor is described. The operation of this technique has been investigated by means of detailed computer simulations. Results of calculations for the geometry and operation conditions of the WASA pellet generator are presented and discussed. We have found that for every given pellet size, there is a set of operation parameters where the efficiency of the WASA hydrogen pellet target operation is considerably increased. Moreover, the results of presented computer simulations clearly show that the future PANDA pellet target setup can be realized with the use of much smaller (and cheaper) vacuum pumps than those used at present in the WASA hydrogen pellet target. To qualitatively improve the PANDA hydrogen pellet target performance we have proposed the use of a novel flow focusing method of Ganan-Calvo and Barreto (1997,1999) combined with the use of conventional vacuum injection capillary. Possibilities of this approach for the PANDA pellet target production have been also explored by means of computer simulations. The results of these simulations show that the use of this new approach looks very promising and in particular, there is no need here to use of expensive ultra-pure hydrogen to prevent nozzle clogging or freezing up due to impurities and it will allow simple, fast, smooth and a wide range of change of pellet sizes in accordance with requirements of different experiments at the PANDA detector. In this article we also propose and describe the idea of a new technique to break up a liquid microjet into microdroplets using a process of liquid jet evaporation under pulsed laser beam irradiation. This technique should be experimentally checked before it may be used in the design of the future PANDA pellet target setup.

  6. Misfire: An Operational Critique of Operation Iraqi Freedom (OIF) Targeting Strategy

    National Research Council Canada - National Science Library

    Martin, John D

    2005-01-01

    ...." However, while PGMs were instrumental in accomplishing the primary national-strategic objective of regime removal in OIF, targeting concentrated exclusively on leadership, command and control (C2...

  7. Feasibility of ''gas target'' mode of divertor operation in ITER

    International Nuclear Information System (INIS)

    Kukushkin, A.S.

    1994-01-01

    Power load upon the divertor target remains one of the most critical issues for a tokamak reactor. Simple estimates, confirmed by 2D modelling, together with some indications from tokamak experiments, showed that the profile of power flow gets narrower along with increase of the reactor power, because strong temperature dependence of the parallel heat conductance, χ parallel αΤ 5/2 , favours parallel heat transport in competition with the cross-field one. This leads to unacceptable peak loads and makes one to look for a means to spread the power more evenly across the magnetic field. The scope of the present paper is to show the results of the modelling studies and to discuss the physical and computational issues which are still missing or are insufficiently developed. I must apologize for partiality for my own calculations with the DDC83 code, but there are some reasons justifying this: they have been the first calculations on this issue, they seem to be the most extensive, and they are certainly the most familiar to me. (orig.)

  8. 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.

  9. 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.

  10. Experimental modelling of plasma-graphite surface interaction in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Martynenko, Yu.V.; Guseva, M.I.; Gureev, V.M.; Danelyan, L.S.; Neumoin, V.E.; Petrov, V.B.; Khripunov, B.I.; Sokolov, Yu.A.; Stativkina, O.V.; Stolyarova, V.G. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation); Vasiliev, V.I.; Strunnikov, V.M. [TRINITI, Troizk (Russian Federation)

    1998-10-01

    The investigation of graphite erosion under normal operation ITER regime and disruption was performed by means of exposure of RGT graphite samples in a stationary deuterium plasma to a dose of 10{sup 22} cm{sup -2} and subsequent irradiation by power (250 MW/cm{sup 2}) pulse deuterium plasma flow imitating disruption. The stationary plasma exposure was carried out in the installation LENTA with the energy of deuterium ions being 200 eV at target temperatures of 770 C and 1150 C. The preliminary exposure in stationary plasma at temperature of physical sputtering does not essentially change the erosion due to a disruption, whereas exposure at the temperature of radiation enhanced sublimation dramatically increases the erosion due to disruption. In the latter case, the depth of erosion due to a disruption is determined by the depth of a layer with decreased strength. (orig.) 9 refs.

  11. 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...

  12. 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.

  13. 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

  14. Evaluation of Radiation Exposure Hazard from Squaw Targets Used in Operations WIGWAM and HARDTACK-1

    Science.gov (United States)

    2016-07-01

    targets during hull-damage testing at the underwater nuclear detonation at Operation WIGWAM, a deep underwater nuclear test conducted as part of the...1945–1962 United States series of atmospheric nuclear tests (DTRA, 2014; USAF, 1955). In preparation for the Operation WIGWAM test, the three...Weary et al., 1981, Figure 1-4) 6 In 1958, SQUAW-29 was used again as a submerged target at a nuclear detonation . This took place at Enewetak

  15. Evaluation of Radiation Exposure Hazard from Squaw Targets Used in Operations WIGWAM and HARDTACK-I

    Science.gov (United States)

    2016-07-01

    targets during hull-damage testing at the underwater nuclear detonation at Operation WIGWAM, a deep underwater nuclear test conducted as part of the...1945–1962 United States series of atmospheric nuclear tests (DTRA, 2014; USAF, 1955). In preparation for the Operation WIGWAM test, the three...Weary et al., 1981, Figure 1-4) 6 In 1958, SQUAW-29 was used again as a submerged target at a nuclear detonation . This took place at Enewetak

  16. 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.)

  17. Monitoring production target thickness

    International Nuclear Information System (INIS)

    Oothoudt, M.A.

    1993-01-01

    Pion and muon production targets at the Clinton P. Anderson Meson Physics Facility consist of rotating graphite wheels. The previous target thickness monitoring Procedure scanned the target across a reduced intensity beam to determine beam center. The fractional loss in current across the centered target gave a measure of target thickness. This procedure, however, required interruption of beam delivery to experiments and frequently indicated a different fractional loss than at normal beam currents. The new monitoring Procedure compares integrated ups and downs toroid current monitor readings. The current monitors are read once per minute and the integral of readings are logged once per eight-hour shift. Changes in the upstream to downstream fractional difference provide a nonintrusive continuous measurement of target thickness under nominal operational conditions. Target scans are now done only when new targets are installed or when unexplained changes in the current monitor data are observed

  18. 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....

  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. 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.

  1. 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

  2. Detection algorithm of infrared small target based on improved SUSAN operator

    Science.gov (United States)

    Liu, Xingmiao; Wang, Shicheng; Zhao, Jing

    2010-10-01

    The methods of detecting small moving targets in infrared image sequences that contain moving nuisance objects and background noise is analyzed in this paper. A novel infrared small target detection algorithm based on improved SUSAN operator is put forward. The algorithm selects double templates for the infrared small target detection: one size is greater than the small target point size and another size is equal to the small target point size. First, the algorithm uses the big template to calculate the USAN of each pixel in the image and detect the small target, the edge of the image and isolated noise pixels; Then the algorithm uses the another template to calculate the USAN of pixels detected in the first step and improves the principles of SUSAN algorithm based on the characteristics of the small target so that the algorithm can only detect small targets and don't sensitive to the edge pixels of the image and isolated noise pixels. So the interference of the edge of the image and isolate noise points are removed and the candidate target points can be identified; At last, the target is detected by utilizing the continuity and consistency of target movement. The experimental results indicate that the improved SUSAN detection algorithm can quickly and effectively detect the infrared small targets.

  3. 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

  4. Generic study on the design and operation of high power targets

    Directory of Open Access Journals (Sweden)

    A. Ahmad

    2014-02-01

    Full Text Available With the move towards beam power in the range of 1–10 MW, a thorough understanding of the response of target materials and auxiliary systems to high power densities and intense radiation fields is required. This paper provides insight into three major aspects related to the design and operation of high power solid targets: thermal stresses, coolant performance, and radiation damage. Where appropriate, a figure-of-merit approach is followed to facilitate the comparison between different target or coolant candidates. The section on radiation damage reports total and spatial variations of displacement-per-atom and helium production levels in different target materials.

  5. Category-based attentional guidance can operate in parallel for multiple target objects.

    Science.gov (United States)

    Jenkins, Michael; Grubert, Anna; Eimer, Martin

    2018-04-30

    The question whether the control of attention during visual search is always feature-based or can also be based on the category of objects remains unresolved. Here, we employed the N2pc component as an on-line marker for target selection processes to compare the efficiency of feature-based and category-based attentional guidance. Two successive displays containing pairs of real-world objects (line drawings of kitchen or clothing items) were separated by a 10 ms SOA. In Experiment 1, target objects were defined by their category. In Experiment 2, one specific visual object served as target (exemplar-based search). On different trials, targets appeared either in one or in both displays, and participants had to report the number of targets (one or two). Target N2pc components were larger and emerged earlier during exemplar-based search than during category-based search, demonstrating the superior efficiency of feature-based attentional guidance. On trials where target objects appeared in both displays, both targets elicited N2pc components that overlapped in time, suggesting that attention was allocated in parallel to these target objects. Critically, this was the case not only in the exemplar-based task, but also when targets were defined by their category. These results demonstrate that attention can be guided by object categories, and that this type of category-based attentional control can operate concurrently for multiple target objects. Copyright © 2018 Elsevier B.V. All rights reserved.

  6. 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)

  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. An Overview of the Target Fabrication Operations at Lawrence Livermore National Laboratory

    International Nuclear Information System (INIS)

    Hibbard, R L; Bono, M J

    2005-01-01

    The Target Engineering team at Lawrence Livermore National Laboratory (LLNL) builds precision laser targets for the National Ignition Facility (NIF) and the Omega Laser in Rochester, NY, and other experimental facilities. The physics requirements demand precision in these targets, which creates a constant need for innovative manufacturing processes. As experimental diagnostics improve, there is greater demand for precision in fabrication, assembly, metrology, and documentation of as-built targets. The team specializes in meso-scale fabrication with core competencies in diamond turning, assembly, and metrology. Figure 1 shows a typical diamond turning center. The team builds over 200 laser targets per year in batches of five to fifteen targets. Thus, all are small-lot custom builds, and most are novel designs requiring engineering and process development. Component materials are metals, polymers and low density aerogel foams. Custom fixturing is used to locate parts on the Diamond Turning Machines (DTM) and assembly stations. This ensures parts can be repeatably located during manufacturing operations. Most target builds involve a series of fabricating one surface with features and then relocating the components on another fixture to finish the opposite side of the component. These components are then assembled to complete multiple-component targets. These targets are typically built one at a time. Cost and efficiency are issues with production of targets, and the team is developing batch processing techniques to meet precision target specifications and cost goals. Three example target builds will highlight some of the fabrication and material issues faced at LLNL. A low temperature Rayleigh Taylor target shows how multiple precision targets can be fabricated out of a single large disk. The ignition double shell targets highlight the required manufacturing complexity. A low density aerogel target highlights some material handling and assembly issues. The metrology

  9. 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.

  10. 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

  11. 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.

  12. Simulation-optimization model of reservoir operation based on target storage curves

    Directory of Open Access Journals (Sweden)

    Hong-bin Fang

    2014-10-01

    Full Text Available This paper proposes a new storage allocation rule based on target storage curves. Joint operating rules are also proposed to solve the operation problems of a multi-reservoir system with joint demands and water transfer-supply projects. The joint operating rules include a water diversion rule to determine the amount of diverted water in a period, a hedging rule based on an aggregated reservoir to determine the total release from the system, and a storage allocation rule to specify the release from each reservoir. A simulation-optimization model was established to optimize the key points of the water diversion curves, the hedging rule curves, and the target storage curves using the improved particle swarm optimization (IPSO algorithm. The multi-reservoir water supply system located in Liaoning Province, China, including a water transfer-supply project, was employed as a case study to verify the effectiveness of the proposed join operating rules and target storage curves. The results indicate that the proposed operating rules are suitable for the complex system. The storage allocation rule based on target storage curves shows an improved performance with regard to system storage distribution.

  13. Proton irradiated graphite grades for a long baseline neutrino facility experiment

    Directory of Open Access Journals (Sweden)

    N. Simos

    2017-07-01

    Full Text Available In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF of the Deep Underground Neutrino Experiment (DUNE four graphite grades were irradiated with protons in the energy range of 140–180 MeV, to peak fluence of ∼6.1×10^{20}  p/cm^{2} and irradiation temperatures between 120–200 °C. The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young’s modulus. The proton fluence level of ∼10^{20}  cm^{−2} where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite

  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. 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

  16. Influence of ion source configuration and its operation parameters on the target sputtering and implantation process.

    Science.gov (United States)

    Shalnov, K V; Kukhta, V R; Uemura, K; Ito, Y

    2012-06-01

    In the work, investigation of the features and operation regimes of sputter enhanced ion-plasma source are presented. The source is based on the target sputtering with the dense plasma formed in the crossed electric and magnetic fields. It allows operation with noble or reactive gases at low pressure discharge regimes, and, the resulting ion beam is the mixture of ions from the working gas and sputtering target. Any conductive material, such as metals, alloys, or compounds, can be used as the sputtering target. Effectiveness of target sputtering process with the plasma was investigated dependently on the gun geometry, plasma parameters, and the target bias voltage. With the applied accelerating voltage from 0 to 20 kV, the source can be operated in regimes of thin film deposition, ion-beam mixing, and ion implantation. Multi-component ion beam implantation was applied to α-Fe, which leads to the surface hardness increasing from 2 GPa in the initial condition up to 3.5 GPa in case of combined N(2)-C implantation. Projected range of the implanted elements is up to 20 nm with the implantation energy 20 keV that was obtained with XPS depth profiling.

  17. Repetitive laser fusion experiment and operation using a target injection system

    International Nuclear Information System (INIS)

    Nishimura, Yasuhiko; Komeda, Osamu; Mori, Yoshitaka

    2017-01-01

    Since 2008, a collaborative research project on laser fusion development based on a high-speed ignition method using repetitive laser has been carried out with several collaborative research institutes. This paper reports the current state of operation of high repetition laser fusion experiments, such as target introduction and control based on a target injection system that allows free falling under 1 Hz, using a high repetition laser driver that has been under research and development, as well as the measurement of targets that freely fall. The HAMA laser driver that enabled high repetition fusion experiments is a titanium sapphire laser using a diode-pumped solid-state laser KURE-I of green light output as a driver pump light source. In order to carry out high repetition laser fusion experiments, the target injection device allows free falling of deuterated polystyrene solid sphere targets of 1 mm in diameter under 1 Hz. The authors integrated the developed laser and injection system, and succeeded first in the world in making the nuclear fusion reaction continuously by hitting the target to be injected with laser, which is essential technology for future laser nuclear fusion reactor. In order to realize repetition laser fusion experiments, stable laser, target synchronization control, and target position measurement technologies are indispensable. (A.O.)

  18. Graphite moderator lifecycle behaviour. Proceedings of a specialists meeting

    International Nuclear Information System (INIS)

    1996-08-01

    The meeting provided the forum for graphite specialists representing operating and research organizations worldwide to exchange information in the following areas: the status of graphite development; operation and safety procedures for existing and future graphite moderated reactors; graphite testing techniques; review of the experiences gained and data acquired on the influence of neutron irradiation and oxidizing conditions on key graphite properties; and to exchange information useful for decommissioning activities. The participants provided twenty-seven papers on behalf of their countries and respective technical organizations. An open discussion followed each of the presentations. A consistently reoccurring theme throughout the specialists meeting was the noticeable reduction in the number of graphite experts remaining the nuclear power industry. Graphite moderated power reactors have provided a significant contribution to the generation of electricity throughout the past forty years and will continue to be a prominent energy source for the future. Yet, many of the renowned experts in the field of nuclear graphites are nearing the end of their careers without apparent replacement. This, coupled with changes in the focus on nuclear power by some industrialized countries, has prompted the IAEA to initiate an evaluation on the feasibility and interest by Member States of establishing a central archive facility for the storage of data on irradiated graphites. Refs, figs, tabs

  19. Graphite moderator lifecycle behaviour. Proceedings of a specialists meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    The meeting provided the forum for graphite specialists representing operating and research organizations worldwide to exchange information in the following areas: the status of graphite development; operation and safety procedures for existing and future graphite moderated reactors; graphite testing techniques; review of the experiences gained and data acquired on the influence of neutron irradiation and oxidizing conditions on key graphite properties; and to exchange information useful for decommissioning activities. The participants provided twenty-seven papers on behalf of their countries and respective technical organizations. An open discussion followed each of the presentations. A consistently reoccurring theme throughout the specialists meeting was the noticeable reduction in the number of graphite experts remaining the nuclear power industry. Graphite moderated power reactors have provided a significant contribution to the generation of electricity throughout the past forty years and will continue to be a prominent energy source for the future. Yet, many of the renowned experts in the field of nuclear graphites are nearing the end of their careers without apparent replacement. This, coupled with changes in the focus on nuclear power by some industrialized countries, has prompted the IAEA to initiate an evaluation on the feasibility and interest by Member States of establishing a central archive facility for the storage of data on irradiated graphites. Refs, figs, tabs.

  20. Target motion predictions for pre-operative planning during needle-based interventions

    NARCIS (Netherlands)

    op den Buijs, J.; Abayazid, Momen; de Korte, Chris L.; Misra, Sarthak

    During biopsies, breast tissue is subjected to displacement upon needle indentation, puncture, and penetration. Thus, accurate needle placement requires pre-operative predictions of the target motions. In this paper, we used ultrasound elastography measurements to non-invasively predict elastic

  1. Operational limit of a planar DC magnetron cluster source due to target erosion

    International Nuclear Information System (INIS)

    Rai, A.; Mutzke, A.; Bandelow, G.; Schneider, R.; Ganeva, M.; Pipa, A.V.; Hippler, R.

    2013-01-01

    The binary collision-based two dimensional SDTrimSP-2D model has been used to simulate the erosion process of a Cu target and its influence on the operational limit of a planar DC magnetron nanocluster source. The density of free metal atoms in the aggregation region influences the cluster formation and cluster intensity during the target lifetime. The density of the free metal atoms in the aggregation region can only be predicted by taking into account (i) the angular distribution of the sputtered flux from the primary target source and (ii) relative downwards shift of the primary source of sputtered atoms during the erosion process. It is shown that the flux of the sputtered atoms smoothly decreases with the target erosion

  2. 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.

  3. 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)

  4. 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.

  5. Surface coating of graphite pebbles for Korean HCCR TBM

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  6. 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.

  7. Surface coating of graphite pebbles for Korean HCCR TBM

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  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. Effects of TEL Confusers on Operator Target Acquisition Performance with SAR Imagery

    Science.gov (United States)

    1998-12-01

    efforts of everyone involved in the project; and Dr. Nick Patton, for providing assistance with several of the mathematical procedures in the document...processing known as the theory of signal detection (TSD) (Gescheider, 1985; Green & Swets, 1966; Macmillan & Creelman , 1991; Wilson, 1992). A TSD...localizations (Hacker & Ratcliff, 1979; Macmillan & Creelman , 1991). The index of bias in a target localization task provides a measure of the operator’s

  10. 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

  11. 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

  12. 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.

  13. Quantifying microstructural dynamics and electrochemical activity of graphite and silicon-graphite lithium ion battery anodes

    Science.gov (United States)

    Pietsch, Patrick; Westhoff, Daniel; Feinauer, Julian; Eller, Jens; Marone, Federica; Stampanoni, Marco; Schmidt, Volker; Wood, Vanessa

    2016-09-01

    Despite numerous studies presenting advances in tomographic imaging and analysis of lithium ion batteries, graphite-based anodes have received little attention. Weak X-ray attenuation of graphite and, as a result, poor contrast between graphite and the other carbon-based components in an electrode pore space renders data analysis challenging. Here we demonstrate operando tomography of weakly attenuating electrodes during electrochemical (de)lithiation. We use propagation-based phase contrast tomography to facilitate the differentiation between weakly attenuating materials and apply digital volume correlation to capture the dynamics of the electrodes during operation. After validating that we can quantify the local electrochemical activity and microstructural changes throughout graphite electrodes, we apply our technique to graphite-silicon composite electrodes. We show that microstructural changes that occur during (de)lithiation of a pure graphite electrode are of the same order of magnitude as spatial inhomogeneities within it, while strain in composite electrodes is locally pronounced and introduces significant microstructural changes.

  14. 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)

  15. Technology strategy for integrated operations and real time reservoir management; Technology Target Areas; TTA5 - Integrated operations and RTRM

    Energy Technology Data Exchange (ETDEWEB)

    2007-07-01

    In Norway Integrated Operations (IO) is a concept which in the first phase (G1) has been used to describe how to integrate processes and people onshore and offshore using ICT solutions and facilities that improve onshore's ability to support offshore operationally. The second generation (G2) Integrated Operations aims to help operators utilize vendors' core competencies and services more efficiently. Utilizing digital services and vendor products, operators will be able to update reservoir models, drilling targets and well trajectories as wells are drilled, manage well completions remotely, optimize production from reservoir to export lines, and implement condition-based maintenance concepts. The total impact on production, recovery rates, costs and safety will be profound. When the international petroleum business moves to the Arctic region the setting is very different from what is the case on the Norwegian Continental Shelf (NCS) and new challenges will arise. The Norwegian Ministry of Environment has recently issued an Integrated Management Plan for the Barents Sea where one focus is on 'Monitoring of the Marine Environment in the North'. The Government aims to establish a new and more coordinated system for monitoring the marine ecosystems in the north. A representative group consisting of the major Operators, the Service Industry, Academia and the Authorities have developed the enclosed strategy for the OG21 Integrated Operations and Real Time Reservoir Management (IO and RTRM) Technology Target Area (TTA). Major technology and work process research and development gaps have been identified in several areas: Bandwidth down-hole to surface; Sensor development including Nano-technology; Cross discipline use of Visualisation, Simulation and model development particularly in Drilling and Reservoir management areas; Software development in terms of data handling, model updating and calculation speed; Enabling reliable and robust communications particularly for

  16. Technology strategy for integrated operations and real time reservoir management; Technology Target Areas; TTA5 - Integrated operations and RTRM

    Energy Technology Data Exchange (ETDEWEB)

    2007-07-01

    In Norway Integrated Operations (IO) is a concept which in the first phase (G1) has been used to describe how to integrate processes and people onshore and offshore using ICT solutions and facilities that improve onshore's ability to support offshore operationally. The second generation (G2) Integrated Operations aims to help operators utilize vendors' core competencies and services more efficiently. Utilizing digital services and vendor products, operators will be able to update reservoir models, drilling targets and well trajectories as wells are drilled, manage well completions remotely, optimize production from reservoir to export lines, and implement condition-based maintenance concepts. The total impact on production, recovery rates, costs and safety will be profound. When the international petroleum business moves to the Arctic region the setting is very different from what is the case on the Norwegian Continental Shelf (NCS) and new challenges will arise. The Norwegian Ministry of Environment has recently issued an Integrated Management Plan for the Barents Sea where one focus is on 'Monitoring of the Marine Environment in the North'. The Government aims to establish a new and more coordinated system for monitoring the marine ecosystems in the north. A representative group consisting of the major Operators, the Service Industry, Academia and the Authorities have developed the enclosed strategy for the OG21 Integrated Operations and Real Time Reservoir Management (IO and RTRM) Technology Target Area (TTA). Major technology and work process research and development gaps have been identified in several areas: Bandwidth down-hole to surface; Sensor development including Nano-technology; Cross discipline use of Visualisation, Simulation and model development particularly in Drilling and Reservoir management areas; Software development in terms of data handling, model updating and calculation speed; Enabling reliable and robust communications

  17. EU 2030 targets 'unachievable' without long-term nuclear operation

    Energy Technology Data Exchange (ETDEWEB)

    Mitev, Lubomir [NucNet, Brussels (Belgium)

    2015-01-15

    Nuclear energy will continue to support greenhouse gas emission reduction targets until 2020, but without decisions on long-term operation of ageing reactors, it will be difficult for the EU to meet its 2030 targets, International Energy Agency (IEA) executive director Maria van der Hoeven, tells NucNet in an interview. The IEA has quite a few remarks and questions related to the EU goals of competitiveness, security of supply and sustainability. It is good to have these targets, but up until now the EU is missing the direct connection between the three goals. What is mostly needed to achieve the goals is to finalise the EU's internal energy market. Secondly cost-effective climate and energy policies are needed because it is not only about climate and energy, but also about economic development and competitiveness. The ageing EU reactor fleet requires country-level and owner/operator-level decisions in the short term regarding plant safety regulations, plant upgrades, uprates, lifetime extensions and licence renewals. Upgrading and uprating existing nuclear plants is one of the cheapest ways of producing carbon-free electricity in the EU. Without long-term operation, the IEA expects nuclear capacity in the EU could fall by a factor of six by 2030 and that will make it more difficult to achieve the EU's 2030 climate targets. Public opinion is an important topic for the acceptance of all energy sources and it is different in all IEA member countries. Europe is very sensitive to almost all forms of energy, including wind turbines and solar panels. This is linked to a lack of information, so we need more and better transparency on information for people.

  18. 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.

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  20. Graphitization of diamond with a metallic coating on ferritic matrix

    International Nuclear Information System (INIS)

    Cabral, Stenio Cavalier; Oliveira, Hellen Cristine Prata de; Filgueira, Marcello

    2010-01-01

    Iron is a strong catalyst of graphitization of diamonds. This graphitization occurs mainly during the processing of composites - conventional sintering or hot pressing, and during cutting operations. Aiming to avoid or minimize this deleterious effect, there is increasing use of diamond coated with metallic materials in the production of diamond tools processed via powder metallurgy. This work studies the influence of Fe on diamond graphitization diamond-coated Ti after mixing of Fe-diamonds, hot pressing parameters were performed with 3 minutes/35MPa/900 deg C - this is the condition of pressing hot used in industry for production of diamond tools. Microstructural features were observed by SEM, diffusion of Fe in diamond was studied by EDS. Graphitization was analyzed by X-ray diffraction and Raman spectroscopy. It was found that Fe not activate graphitization on the diamond under the conditions of hot pressing. (author)

  1. Impermeable Graphite: A New Development for Embedding Radioactive Waste

    International Nuclear Information System (INIS)

    Fachinger, Johannes

    2016-01-01

    Irradiated graphite has to be handled as radioactive waste after the operational period of the reactor. However, the waste management of irradiated graphite e.g. from the Spanish Vandellos reactor shows, that waste management of even low contaminated graphite could be expensive and requires special retrieval, treatment and disposal technologies for safe long term storage as low or medium radioactive waste. FNAG has developed an impermeable graphite matrix (IGM) as nuclear waste embedding material. This IGM provides a long term stable enclosure of radioactive waste and can reuse irradiated graphite as feedstock material. Therefore, no additional disposal volume is required if e.g. concrete waste packages were replaced by IGM waste packages. The variability of IGM as embedding has been summarized in the following paper usable for metal scraps, ion exchange resins or debris from buildings. Furthermore the main physical, chemical and structural properties are described. (author)

  2. European DEMO divertor target: Operational requirements and material-design interface

    Directory of Open Access Journals (Sweden)

    J.H. You

    2016-12-01

    Full Text Available Recently, an integrated program of conceptual design activities for the European DEMO reactor was launched in the framework of the EUROfusion Consortium, where reliable power handling capability was identified as one of the most critical scientific as well as technological challenges for a DEMO reactor. The divertor is the key in-vessel plasma-facing component being in charge of power exhaust and removal of impurity particles. The DEMO divertor target will have to withstand extreme thermal loads where the local peak heat flux is expected to reach up to 20 MW/m2 during slow transient events in DEMO. To assure sufficient heat removal capability of the divertor target against normal and transient operational scenarios under expected cumulative neutron dose of up to 13 dpa is one of the fundamental engineering challenges imposed on target design. To develop the design of the DEMO divertor and related technologies, an R&D work package ‘Divertor’ has been set up in this consortium. The subproject ‘Target Development’ is devoted to the development of the conceptual design and the core technologies of the plasma-facing target. Devising and implementing novel structural heat sink materials (e.g. W/Cu composites to advanced target design concepts is one of the major objectives of this subproject. In this paper, the underlying design requirements imposed by the envisaged power exhaust goal and the prominent material-design interface issues are discussed. In addition, the candidate design concepts being currently considered are presented together with the related material issues. Finally, the first results achieved so far are presented.

  3. Technical development of graphite waste treatment in NUPEC

    International Nuclear Information System (INIS)

    Saishu, S.; Inoue, T.

    2001-01-01

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

  4. Electronic properties of graphite

    International Nuclear Information System (INIS)

    Schneider, J.

    2010-10-01

    In this thesis, low-temperature magneto-transport (T ∼ 10 mK) and the de Haas-van Alphen effect of both natural graphite and highly oriented pyrolytic graphite (HOPG) are examined. In the first part, low field magneto-transport up to B = 11 T is discussed. A Fourier analysis of the background removed signal shows that the electric transport in graphite is governed by two types of charge carriers, electrons and holes. Their phase and frequency values are in agreement with the predictions of the SWM-model. The SWM-model is confirmed by detailed band structure calculations using the magnetic field Hamiltonian of graphite. The movement of the Fermi at B > 2 T is calculated self-consistently assuming that the sum of the electron and hole concentrations is constant. The second part of the thesis deals with high field magneto-transport of natural graphite in the magnetic field range 0 ≤ B ≤ 28 T. Both spin splitting of magneto-transport features in tilted field configuration and the onset of the charge density wave (CDW) phase for different temperatures with the magnetic field applied normal to the sample plane are discussed. Concerning the Zeeman effect, the SWM calculations including the Fermi energy movement require a g-factor of g* equal to 2.5 ± 0.1 to reproduce the spin spilt features. The measurements of the charge density wave state confirm that its onset magnetic field can be described by a Bardeen-Cooper-Schrieffer (BCS)-type formula. The measurements of the de Haas-van Alphen effect are in agreement with the results of the magneto-transport measurements at low field. (author)

  5. HTCAP-1: a program for calcuating operating temperatures in HFIR target irradiation experiments

    International Nuclear Information System (INIS)

    Kania, M.J.; Howard, A.M.

    1980-06-01

    The thermal modeling code, HTCAP-1, calculates in-reactor operating temperatures of fueled specimens contained in the High Flux Isotope Reactor (HFIR) target irradiation experiments (HT-series). Temperature calculations are made for loose particle and bonded fuel rod specimens. Maximum particle surface temperatures are calculated for the loose particles and centerline and surface temperatures for the fuel rods. Three computational models are employed to determine fission heat generation rates, capsule heat transfer analysis, and specimen temperatures. This report is also intended to be a users' manual, and the application of HTCAP-1 to the HT-34 irradiation capsule is presented

  6. Radiation damage and life-time evaluation of RBMK graphite stack

    Energy Technology Data Exchange (ETDEWEB)

    Platonov, P A; Chugunov, O K; Manevsky, V N; Karpukhin, V I [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation). Reactor Material Div.

    1996-08-01

    At the present time there are 11 NPP units with RBMK reactors in operation in Russia, with the oldest now in operation 22 years. Design life-time of the RBMK-1000 reactor is 30 years. This paper addresses the evaluation of RBMK graphite stack life-time. It is the practice in Russia to evaluate the reliability of the channel reactor graphite stack using at least three criteria: degradation of physical-mechanical properties of graphite, preservation of the graphite brick integrity, and degradation of the graphite stack as a structure. Stack life-time evaluation by different criteria indicates that the most realistic approach may be realized on the basis of the criteria of brick cracking and degradation of the graphite stack as a structure. The RBMK reactor graphite stack life-time depends on its temperature and for different units it may be different. (author). 2 refs, 10 figs.

  7. The Holistic Targeting (HOT) methodology as the means to improve Information Operations (IO) target development and prioritization

    OpenAIRE

    Ieva, Christopher S.

    2008-01-01

    Prioritization. In response to this challenge, this study proposes five recommendations to enhance IO integration into the Joint Targeting Cycle: the use of interim IO Joint Munitions Effectiveness Manual (JMEM) techniques to better forecast cognitive effects, the adoption of the Measure of Worth (MOW) model to assess IO effects, the HOT methodology to develop and prioritize IO targets, the use of compendium software facilitate targeting problem understanding and the network analysis to...

  8. 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

  9. Strategy for Handling and Treatment of INPP RBMK-1500 Irradiated Graphite

    International Nuclear Information System (INIS)

    Oryšaka, A.

    2016-01-01

    There are two RBMK-1500 water-cooled graphite-moderated channel-type power reactors at Ignalina NPP. After the final shutdown of the INPP, radioactive i-graphite dismantling, handling, conditioning, storage and disposal is an important part of the decommissioning activities. The core of the INPP unit 1 and 2 contains about 3600 tons of i-graphite. Formation of activation products strongly depends on the contents of impurities, operational mode and concentration of impurities in the graphite. The case study for INPP envisages the analysis of possibilities of graphite handling and treatment in the context of immediate decommissioning. (author)

  10. Operational Benefits of Meeting California's Energy Storage Targets

    Energy Technology Data Exchange (ETDEWEB)

    Eichman, Josh; Denholm, Paul; Jorgenson, Jennie; Helman, Udi

    2016-05-01

    In October 2013, the California Public Utilities Commission (CPUC) finalized procurement targets and other requirements to its jurisdictional utilities for a minimum of 1,325 MW of 'viable and cost-effective' energy storage systems by 2020. The goal of this study is to explore several aspects of grid operations in California and the Western Interconnection resulting from meeting the CPUC storage targets. We perform this analysis using a set of databases and grid simulation tools developed and implemented by the CPUC, the California Independent System Operator (CAISO), and the California Energy Commission (CEC) for the CPUC's Long-term Procurement Plan (LTPP). The 2014 version of this database contains information about generators, storage, transmission, and electrical demand, for California in the year 2024 for both 33 percent and 40 percent renewable energy portfolios. We examine the value of various services provided by energy storage in these scenarios. Sensitivities were performed relating to the services energy storage can provide, the capacity and duration of storage devices, export limitations, and negative price floor variations. Results show that a storage portfolio, as outlined by the CPUC, can reduce curtailment and system-wide production costs for 33 percent and 40 percent renewable scenarios.

  11. Evaluation of Accelerated Graphitic Corrosion Test of Gray Cast Iron

    International Nuclear Information System (INIS)

    Kim, Jeong Hyeon; Hong, Jong Dae; Chang Heui; Na, Kyung Hwan; Lee, Jae Gon

    2011-01-01

    In operating nuclear power plants, gray cast iron is commonly used as materials for various non-safety system components including pipes in fire water system, valve bodies, bonnets, and pump castings. In such locations, operating condition does not require alloy steels with excellent mechanical properties. But, a few corrosion related degradation, or graphitic corrosion is frequently occurred to gray cast iron during the long-term operation in nuclear power plant. Graphitic corrosion is selective leaching of iron from gray cast iron, where iron gets removed and graphite grains remain intact. In U.S.A., one-time visual inspection and hardness measurement are required from regulatory body to detect the graphitic corrosion for the life extension evaluation of the operating nuclear power plant. In this study, experiments were conducted to make accelerated graphitic corrosion of gray cast iron using electrochemical method, and hardness was measured for the specimens to establish the correlation between degree of graphitic corrosion and surface hardness of gray cast iron

  12. 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.

  13. 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

  14. 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.

  15. 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.

  16. 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.

  17. A universal molecular translator for non-nucleic acid targets that enables dynamic DNA assemblies and logic operations.

    Science.gov (United States)

    Tang, Wei; Hu, Shichao; Wang, Huaming; Zhao, Yan; Li, Na; Liu, Feng

    2014-11-28

    A universal molecular translator based on the target-triggered DNA strand displacement was developed, which was able to convert various kinds of non-nucleic acid targets into a unique output DNA. This translation strategy was successfully applied in directing dynamic DNA assemblies and in realizing three-input logic gate operations.

  18. Carbon-14 in neutron-irradiated graphite for graphite-moderated reactors. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, Kimio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Matsuo, Hideto [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokyo (Japan)

    2002-12-01

    The graphite moderated gas cooled reactor operated by the Japan Atomic Power Company was stopped its commercial operation on March 1998, and the decommissioning process has been started. Graphite material is often used as the moderator and the reflector materials in the core of the gas cooled reactor. During the operation, a long life nuclide of {sup 14}C is generated in the graphite by several transmutation reactions. Separation of {sup 14}C isotope and the development of the separation method have been recognized to be critical issues for the decommissioning of the reactor core. To understand the current methodologies for the carbon isotope separation, literature on the subject was surveyed. Also, those on the physical and chemical behavior of {sup 14}C were surveyed. This is because the larger part of the nuclides in the graphite is produced from {sup 14}N by (n,p) reaction, and the location of them in the material tends to be different from those of the other carbon atoms. This report summarizes the result of survey on the open literature about the behavior of {sup 14}C and the separation methods, including the list of the literature on these subjects. (author)

  19. Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

    International Nuclear Information System (INIS)

    Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

    2005-01-01

    G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  20. Atomic resolution images of graphite in air

    Energy Technology Data Exchange (ETDEWEB)

    Grigg, D.A.; Shedd, G.M.; Griffis, D.; Russell, P.E.

    1988-12-01

    One sample used for proof of operation for atomic resolution in STM is highly oriented pyrolytic graphite (HOPG). This sample has been imaged with many different STM`s obtaining similar results. Atomic resolution images of HOPG have now been obtained using an STM designed and built at the Precision Engineering Center. This paper discusses the theoretical predictions and experimental results obtained in imaging of HOPG.

  1. 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.

  2. Target Bosnia: Integrating Information Activities in Peace Operations. NATO-Led Operations in Bosnia-Herzegovina, December 1995-1997

    National Research Council Canada - National Science Library

    Siegel, Pascale C

    1998-01-01

    This monograph examines the role of information in Peace Support Operations (PSO) and its impact on command and control through the prism of NATO-led operations in Bosnia-Herzegovina from December 1995 into 1997...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-02-15

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

  4. Energy Storage System Control Algorithm by Operating Target Power to Improve Energy Sustainability of Smart Home

    Directory of Open Access Journals (Sweden)

    Byeongkwan Kang

    2018-01-01

    Full Text Available As energy issues are emerging around the world, a variety of smart home technologies aimed at realizing zero energy houses are being introduced. Energy storage system (ESS for smart home energy independence is increasingly gaining interest. However, limitations exist in that most of them are controlled according to time schedules or used in conjunction with photovoltaic (PV generation systems. In consideration of load usage patterns and PV generation of smart home, this study proposes an ESS control algorithm that uses constant energy of energy network while making maximum use of ESS. Constant energy means that the load consumes a certain amount of power under all conditions, which translates to low variability. The proposed algorithm makes a smart home a load of energy network with low uncertainty and complexity. The simulation results show that the optimal ESS operating target power not only makes the smart home use power constantly from the energy network, but also maximizes utilization of the ESS. In addition, since the smart home is a load that uses constant energy, it has the advantage of being able to operate an efficient energy network from the viewpoint of energy providers.

  5. Operational Benefits of Meeting California's Energy Storage Targets

    Energy Technology Data Exchange (ETDEWEB)

    Eichman, Josh [National Renewable Energy Lab. (NREL), Golden, CO (United States); Denholm, Paul [National Renewable Energy Lab. (NREL), Golden, CO (United States); Jorgenson, Jennie [National Renewable Energy Lab. (NREL), Golden, CO (United States); Helman, Udi [Helman Analytics, San Francisco, CA (United States)

    2015-12-18

    In October 2013, the California Public Utilities Commission (CPUC) finalized procurement targets and other requirements to its jurisdictional utilities for a minimum of 1,325 MW of 'viable and cost-effective' energy storage systems by 2020. The goal of this study is to explore several aspects of grid operations in California and the Western Interconnection resulting from meeting the CPUC storage targets. We perform this analysis using a set of databases and grid simulation tools developed and implemented by the CPUC, the California Independent System Operator (CAISO), and the California Energy Commission (CEC) for the CPUC's Long-term Procurement Plan (LTPP). The 2014 version of this database contains information about generators, storage, transmission, and electrical demand, for California in the year 2024 for both 33% and 40% renewable energy portfolios. We examine the value of various services provided by energy storage in these scenarios. Sensitivities were performed relating to the services energy storage can provide, the capacity and duration of storage devices, export limitations, and negative price floor variations. Results show that a storage portfolio, as outlined by the CPUC, can reduce curtailment and system-wide production costs for 33% and 40% renewable scenarios. A storage device that can participate in energy and ancillary service markets provides the grid with the greatest benefit; the mandated storage requirement of 1,325 MW was estimated to reduce the total cost of production by about 78 million per year in the 33% scenario and 144 million per year in the 40% scenario. Much of this value is derived from the avoided start and stop costs of thermal generators and provision of ancillary services. A device on the 2024 California grid and participating in only ancillary service markets can provide the system with over 90% of the value as the energy and ancillary service device. The analysis points to the challenge of new storage

  6. 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.

  7. 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

  8. 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...

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  10. 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.

  11. 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

  12. Structural performance of a graphite blanket in fusion reactors

    International Nuclear Information System (INIS)

    Wolfer, W.G.; Watson, R.D.

    1978-01-01

    Irradiation of graphite in a fusion reactor causes dimensional changes, enhanced creep, and changes in elastic properties and fracture strength. Temperature and flux gradients through the graphite blanket structure produce differential distortions and stress gradients. An inelastic stress analysis procedure is described which treats these variations of the graphite properties in a consistent manner as dictated by physical models for the radiation effects. Furthermore, the procedure follows the evolution of the stress and fracture strength distributions during the reactor operation as well as for possible shutdowns at any time. The lifetime of the graphite structure can be determined based on the failure criterion that the stress at any location exceeds one-half of the fracture strength. This procedure is applied to the most critical component of the blanket module in the SOLASE design

  13. Thermal strain measurements in graphite using electronic speckle pattern interferometry

    International Nuclear Information System (INIS)

    Tamulevicius, S.; Augulis, L.; Augulis, R.; Zabarskas, V.; Levinskas, R.; Poskas, P.

    2001-01-01

    Two 1500 MW(e) RBMK Units are operated at Ignalina NPP in Lithuania. Due to recent decision of the Parliament on the earlier closure of Unit 1, preparatory work for decommissioning have been initiated. Preferred decommissioning strategy is based on delayed dismantling after rather long safe enclosure period. Since graphite is one of the basic and probably the most voluminous components of the reactor internals, a sufficient information on status and behaviour of graphite moderator and reflector during long time safe enclosure period is of special significance. In this context, thermal strain in graphite is one of the parameters requiring particular interest. Electronic speckle pattern interferometry has been proposed and successfully tested to control this parameter using the real samples of graphite from Ignalina NPP Units. (author)

  14. A comparison of predicted and measured graphite moderator behaviour during 16 years' operation of the Windscale advanced gas-cooled reactor

    International Nuclear Information System (INIS)

    Prince, N.

    1980-01-01

    The Windscale AGR has operated since January 1963 at a cumulative load factor of nearly 70% during which time the peak irradiation damage dose has built up to more than 5x10 n/cm 2 (equivalent DIDO nickel), well beyond its original design life. This paper recounts the findings of monitoring measurements on the moderator at high exposure levels with regard to radiolytic oxidation and various aspects of dimensional change behaviour. It is shown that the measured dimensional changes are in good agreement with predictions based on small specimens irradiated in MTR's, thus confirming the absence of any size effect and adding confidence to predictive methods. However, recent measurements of channel straightness show that the observed distortions are only about 10% of the maximum predictions, perhaps due to localised creep at the brick ends creating flats which impart some stability to the columns of moderator bricks. The magnitude of radiolytic oxidation determined by trepanning specimens from the core in 1976 was found to be only about 5%, whereas it was thought possible that peak weight losses would conceivably be as high as 11% due to the depleted concentration of methane inhibitor reaching the brick interior by diffusion processes. It has subsequently been shown by calculation that this result is consistent with the existence of radial pressure drops across the moderator brick walls giving greater penetration of methane inhibitor. (author)

  15. Low cycle fatigue behavior of ITER-like divertor target under DEMO-relevant operation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Li, Muyuan; Werner, Ewald [Lehrstuhl für Werkstoffkunde und Werkstoffmechanik, Technische Universität München, Boltzmannstr. 15, 85748 Garching (Germany); You, Jeong-Ha, E-mail: you@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

    2015-01-15

    Highlights: • LCF behavior of the cooling tube and the interlayer of an ITER-like divertor target is studied. • For the cooling tube, LCF failure will not be an issue under an HHF load of up to 18 MW/m{sup 2}. • Plastic strain in the interlayer is concentrated at the free surface edge of the bond interface. • The predicted LCF lifetime of the interlayer may not meet the design requirement. - Abstract: In this work the low cycle fatigue (LCF) behavior of the copper alloy cooling tube and the copper interlayer of an ITER-like divertor target is reported for nine different combinations of loading and cooling conditions relevant to DEMO divertor operation. The LCF lifetime is presented as a function of loading and cooling conditions considered here by means of cyclic plasticity simulation and using LCF data of materials relevant for ITER. The numerical predictions indicate, that fatigue failure will not be an issue for the copper alloy tube under a high heat flux (HHF) load of up to 18 MW/m{sup 2} as long as it preserves its initial strength. In contrast, the copper interlayer exhibits significant plastic dissipation at the free surface edge of the bond interface adjacent to the cooling tube, where the LCF lifetime is predicted to be below 3000 load cycles for HHF loads higher than 15 MW/m{sup 2}. Most of the bulk region of the copper interlayer away from the free surface edge does not experience severe plastic fatigue and hence does not pose any critical concern as the LCF lifetime is predicted to be at least 7000 load cycles. LCF lifetime decreases as HHF load is increased or coolant temperature is decreased.

  16. 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

  17. 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)

  18. Late-time particle emission from laser-produced graphite plasma

    Energy Technology Data Exchange (ETDEWEB)

    Harilal, S. S.; Hassanein, A.; Polek, M. [School of Nuclear Engineering, Center for Materials Under Extreme Environment, Purdue University, West Lafayette, Indiana 47907 (United States)

    2011-09-01

    We report a late-time ''fireworks-like'' particle emission from laser-produced graphite plasma during its evolution. Plasmas were produced using graphite targets excited with 1064 nm Nd: yttrium aluminum garnet (YAG) laser in vacuum. The time evolution of graphite plasma was investigated using fast gated imaging and visible emission spectroscopy. The emission dynamics of plasma is rapidly changing with time and the delayed firework-like emission from the graphite target followed a black-body curve. Our studies indicated that such firework-like emission is strongly depended on target material properties and explained due to material spallation caused by overheating the trapped gases through thermal diffusion along the layer structures of graphite.

  19. Late-time particle emission from laser-produced graphite plasma

    International Nuclear Information System (INIS)

    Harilal, S. S.; Hassanein, A.; Polek, M.

    2011-01-01

    We report a late-time ''fireworks-like'' particle emission from laser-produced graphite plasma during its evolution. Plasmas were produced using graphite targets excited with 1064 nm Nd: yttrium aluminum garnet (YAG) laser in vacuum. The time evolution of graphite plasma was investigated using fast gated imaging and visible emission spectroscopy. The emission dynamics of plasma is rapidly changing with time and the delayed firework-like emission from the graphite target followed a black-body curve. Our studies indicated that such firework-like emission is strongly depended on target material properties and explained due to material spallation caused by overheating the trapped gases through thermal diffusion along the layer structures of graphite.

  20. R and D and maintenance on graphite tile of divertor region at EAST

    International Nuclear Information System (INIS)

    Ji, X.; Song, Y.T.; Wu, S.T.; Hao, J.; Du, S.; Peng, Y.; Cao, L.; Wang, S.

    2012-01-01

    Highlights: ► Find out the reason of damage of graphite tile. ► Simulation the halo current. ► Stress analysis of graphite tile by ANSYS. ► Do the experiments to test the strength of graphite tile. ► Do the optimization and maintenance of graphite tile. - Abstract: EAST, with full superconducting magnetic coils, had been designed and constructed to address the scientific and engineering issues under steady state operation. The in-vessel components are full graphite tiles as first wall had been operated successfully. In the experiment campaign of 2010, the H mode operation and 1 MA operation have been gotten on EAST. However, in some case, some of the graphite tiles of divertor region are damaged with the plasma parameter enhanced. As most of the damaged graphite tiles are in the divertor region, they are probably damaged by the electro-magnetic force of the halo current when the VDEs occur. The force of the halo current is re-estimated. The structure analysis has been done by the ANSYS software. From the analysis result. It can be obtained that the stress is larger than the allowable stress when the halo current on the graphite tile is larger than 2.7 kA. The tensile testing of the graphite also has been done. As the result, the graphite tiles are damaged when the forces are up to 2400 N. To deal with the problem, two proposes are accepted. In the one hand, the new type graphite material is used, whose tensile strength is up to 45 MPa. In the other hand, the structure of the graphite tiles is optimized.

  1. Target values for nuclear material safeguards measurements - motivation or burden to operators?

    International Nuclear Information System (INIS)

    Weh, R.; Kuhn, K.D.

    1989-01-01

    The analytical determination of material streams and inventories plays an important part in those nuclear facilities called bulk-handling facilities in safeguards terminology. Reprocessing plants and mixed-oxide fabrication facilities are typical examples. With respect to their safeguards, the relevant regulations attach fundamental importance to material accountancy. The balance itself is examined by International Atomic Energy Agency (IAEA) inspectors and within the boundaries of the European Communities by Euratom inspectors as well, with regard to formal correctness. The analytical methods of accountancy in, for example, reprocessing plants, make high demands on the qualifications of the analyst. A conscientious analyst will, of course, try to fulfill his task as well and effectively as possible. These target values will become a burden, however, when they have been drawn up for purely scientific interest and the operator has been urged to achieve them on the pretext of improving safeguards. There are basically two reasons for which the authors have misgivings in this respect. First, the measurement system, which the material balance is based on, has to conform to the latest international regulations. This could easily lead to a permanent obligation of updating for the plants concerned. Second, the goal quantities set by the IAEA will induce an attempt to adjust measurement techniques and chemical analysis to goals totally unsuitable for large-scale plants

  2. Strength degradation of oxidized graphite support column in VHTR

    International Nuclear Information System (INIS)

    Park, Byung Ha; No, Hee Cheon

    2010-01-01

    Air-ingress events caused by large pipe breaks are important accidents considered in the design of Very High Temperature Gas-Cooled Reactors (VHTRs). A main safety concern for this type of event is the possibility of core collapse following the failure of the graphite support column, which can be oxidized by ingressed air. In this study, the main target is to predict the strength of the oxidized graphite support column. Through compression tests for fresh and oxidized graphite columns, the compressive strength of IG-110 was obtained. The buckling strength of the IG-110 column is expressed using the following empirical straight-line formula: σ cr,buckling =91.34-1.01(L/r). Graphite oxidation in Zone 1 is volume reaction and that in Zone 3 is surface reaction. We notice that the ultimate strength of the graphite column oxidized in Zones 1 and 3 only depends on the slenderness ratio and bulk density. Its strength degradation oxidized in Zone 1 is expressed in the following nondimensional form: σ/σ 0 =exp(-kd), k=0.114. We found that the strength degradation of a graphite column, oxidized in Zone 3, follows the above buckling empirical formula as the slenderness of the column changes. (author)

  3. 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

  4. 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

  5. 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)

  6. 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.

  7. Targeted neurosurgical outreach: 5-year follow-up of operative skill transfer and sustainable care in Lima, Peru.

    Science.gov (United States)

    Duenas, Vincent J; Hahn, Edward J; Aryan, Henry E; Levy, Michael V; Jandial, Rahul

    2012-08-01

    This study evaluates the efficacy of operative skill transfer in the context of targeted pediatric outreach missions. In addition, the ability to implement surgical care improvements that are sustainable is investigated. Three 1-week targeted neurosurgical missions were performed (2004-2006) to teach neuroendoscopy, which included donation of the necessary equipment so newly acquired surgical skills could be performed by local neurosurgeons in between and after the departure of the mission team. After the targeted missions were completed, 5 years of neuroendoscopy case follow-up data were obtained. After performing pediatric neurosurgery missions in 2004-2006, with a focus on teaching neuroendoscopy, the host team demonstrated the sustainability of our didactic efforts in the subsequent 5 years by performing cases independently for their citizens. To date, a total of 196 operations have been performed in the past 5 years independent of any visiting team. Effective operative skill transfer to host neurosurgeons can be accomplished with limited international team visits utilizing a targeted approach that minimizes expenditures on personnel and capital. With the priority being teaching of an operative technique, as opposed to perennially performing operations by the mission team, sustainable surgical care was achieved after missions officially concluded.

  8. Statistical considerations of graphite strength for assessing design allowable stresses

    International Nuclear Information System (INIS)

    Ishihara, M.; Mogi, H.; Ioka, I.; Arai, T.; Oku, T.

    1987-01-01

    Several aspects of statistics need to be considered to determine design allowable stresses for graphite structures. These include: 1) Statistical variation of graphite material strength. 2) Uncertainty of calculated stress. 3) Reliability (survival probability) required from operational and safety performance of graphite structures. This paper deals with some statistical considerations of structural graphite for assessing design allowable stress. Firstly, probability distribution functions of tensile and compressive strengths are investigated on experimental Very High Temperature candidated graphites. Normal, logarithmic normal and Weibull distribution functions are compared in terms of coefficient of correlation to measured strength data. This leads to the adaptation of normal distribution function. Then, the relation between factor of safety and fracture probability is discussed on the following items: 1) As the graphite strength is more variable than metalic material's strength, the effect of strength variation to the fracture probability is evaluated. 2) Fracture probability depending on survival probability of 99 ∼ 99.9 (%) with confidence level of 90 ∼ 95 (%) is discussed. 3) As the material properties used in the design analysis are usually the mean values of their variation, the additional effect of these variations on the fracture probability is discussed. Finally, the way to assure the minimum ultimate strength with required survival probability with confidence level is discussed in view of statistical treatment of the strength data from varying sample numbers in a material acceptance test. (author)

  9. Porosity effects in the neutron total cross section of graphite

    International Nuclear Information System (INIS)

    Santisteban, J. R; Dawidowski, J; Petriw, S. N

    2009-01-01

    Graphite has been used in nuclear reactors since the birth of the nuclear industry due to its good performance as a neutron moderator material. Graphite is still an option as moderator for generation IV reactors due to its good mechanical and thermal properties at high operation temperatures. So, there has been renewed interest in a revision of the computer libraries used to describe the neutron cross section of graphite. For sub-thermal neutron energies, polycrystalline graphite shows a larger total cross section (between 4 and 8 barns) than predicted by existing theoretical models (0.2 barns). In order to investigate the origin of this discrepancy we measured the total cross section of graphite samples of three different origins, in the energy range from 0.001 eV to 10 eV. Different experimental arrangements and sample treatments were explored, to identify the effect of various experimental parameters on the total cross section measurement. The experiments showed that the increase in total cross section is due to neutrons scattered around the forward direction. We associate these small-angle scattered neutrons (SANS) to the porous structure of graphite, and formulate a very simple model to compute its contribution to the total cross section of the material. This results in an analytic expression that explicitly depends on the density and mean size of the pores, which can be easily incorporated in nuclear library codes. [es

  10. 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

  11. 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

  12. 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)

  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. 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.

  15. 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

    fraction of the chlorine is released from the graphite during irradiation, both in natural precursor and activated form. The release rate of inactive chlorine and of chlorine-36 was modelled, to allow the calculation of a residual chlorine-36 inventory in operational and decommissioning wastes; actual measurements on operational wastes have confirmed this release. This has subsequently been used to assess the effect on disposal risks. This paper provides a description of the work undertaken by Nirex to establish the residual inventory of chlorine-36 in irradiated nuclear graphite, and outline the implications for disposal of these wastes. (author)

  16. 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...

  17. 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...

  18. 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

  19. 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)

  20. Mobile operators have set ambitious targets – Is it possible to boost network capacity while reducing its energy consumption?

    DEFF Research Database (Denmark)

    Micallef, Gilbert; Mogensen, Preben; Scheck, Hans-Otto

    2012-01-01

    While operators have to upgrade the capacity of their networks, they have committed themselves to reduce their CO2 emissions, partly by reducing their energy consumption. This article investigates the challenges faced by operators and quantifies, through a number of case studies, the impact...... the possible savings by adopting an energy-efficient capacity evolution together with an equipment replacement and site upgrade strategy. Results show that network operators can get relatively close to their targets, with energy reductions of up to 40% noted. While this can be improved further through software...... based energy saving features, further CO2 emissions can be offset through carbon-neutral energy sources....

  1. Safety-technical lay-out of the operational environment of a high-power spallation target system of the megawatt class with mercury as target material

    International Nuclear Information System (INIS)

    Butzek, M.

    2005-06-01

    This thesis is concerning the safety relevant layout of the environment of a mercury based 5-Megawatt-spallation target. All safety relevant aspects related to construction, operation and dismantling as well as economical issues were taken into account. Safety concerns are basically driven by the toxic and radioactive inventory as well as the kind and intensity of radiation produced by the spallation process. Due to significant differences in inventory and radiation between a spallation source and a fission reactor, for the design of the spallation source mentioned above the safety philosophy of a fission reactor must not be used unchanged. Rather than this a systematic study of all safety related boundary conditions is necessary. Within this thesis all safety relevant boundary conditions for this specific type of machine are given. Beside the spatial distribution of different areas inside the target station, influence of medias to be used as well as arising radiation and handling requirements are discussed in detail. A general layout of the target station is presented, serving as a basis for all further component and system development. An enclosure concept for the target station was developed, taking into account the safety relevant issues concerning the mercury used as target materials, the water cooling loops containing massive amounts of tritium as well as the materials used for the moderators potentially forming explosive mixtures. Concept and detailed technical layout of the enclosure system was chosen to guarantee safe operation of the source as well as taking care of requirement arising for handling needs. For design of the shielding different suitable materials have been discussed. A design for assembling the shielding is shown taking into account the safety relevant requirements during operation as well as during dismantling. The neutron beam shutters, buried inside the shielding were designed to optimize handling and positioning issued of the inner part

  2. 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)

  3. 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.

  4. Commissioning and Operation of a Cryogenic Target at HI γS

    Science.gov (United States)

    Kendellen, David; Compton@HIγS Collaboration

    2017-01-01

    We have developed a cryogenic target for use at the High Intensity γ-ray Source (HI γS). The target system is able to liquefy helium-4 (LHe) at 4 K, hydrogen (LH2) at 20 K, or deuterium (LD2) at 23 K to fill a 0.3 L Kapton cell. Liquid temperatures and condenser pressures are recorded throughout each run in order to ensure that the target's areal density is known to 1%. A low-temperature valve enables cycling between full and empty modes in less than 15 minutes. The target is being utilized in a series of experiments which probe the electromagnetic polarizabilities of the nucleon by Compton scattering high-energy photons from the liquid and detecting them with the HI γS NaI Detector Array (HINDA). During a 50-hour-long commissioning run, the target held LHe at 3.17 K, followed by 600 hours of production running with LD2 at 23.9 K. The design of the target will be presented and its performance during these runs will be discussed. Work supported by US Department of Energy contracts DE-FG02-97ER41033, DE-FG02-06ER41422, and DE-SCOO0536

  5. Carbowaste: treatment and disposal of irradiated graphite and other carbonaceous waste

    International Nuclear Information System (INIS)

    Von Lensa, W.; Rizzato, C.; Baginski, K.; Banford, A.W.; Bradbury, D.; Goodwin, J.; Grambow, B.; Grave, M.J.; Jones, A.N.; Laurent, G.; Pina, G.; Vulpius, D.

    2014-01-01

    The European Project on 'Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste (CARBOWASTE)' addressed the retrieval, characterization, treatment, reuse and disposal of irradiated graphite with the following main results: - I-graphite waste features significantly depend on the specific manufacture process, on the operational conditions in the nuclear reactor (neutron dose, atmosphere, temperature etc.) and on radiolytic oxidation leading to partial releases of activation products and precursors during operation. - The neutron activation process generates significant recoil energies breaking pre-existing chemical bonds resulting in dislocations of activation products and new chemical compounds. - Most activation products exist in different chemical forms and at different locations. - I-graphite can be partly purified by thermal and chemical treatment processes leaving more leach-resistant waste products. - Leach tests and preliminary performance analyses show that i-graphite can be safely disposed of in a wide range of disposal systems, after appropriate treatment and/or conditioning. (authors)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  7. Corrosion behavior of a positive graphite electrode in vanadium redox flow battery

    International Nuclear Information System (INIS)

    Liu Huijun; Xu Qian; Yan Chuanwei; Qiao Yonglian

    2011-01-01

    Graphical abstract: The overpotential for gas evolution on positive graphite electrode decreases due to the functional groups of COOH and C=O introduced on the surface of graphite electrode during corrosion process, which can self-catalyze the oxidation of carbon atoms therefore, accelerates corrosion process. Highlights: → Initial potential for gas evolution is higher than 1.60 V vs SCE. → Factors affecting the graphite corrosion are investigated. → Functional groups of COOH and C=O introduced during corrosion process. → The groups can self-catalyze the oxidation of carbon atoms. - Abstract: The graphite plate is easily suffered from corosion because of CO 2 evolution when it acts as the positive electrode for vanadium redox flow battery. The aim is to obtain the initial potential for gas evolution on a positive graphite electrode in 2 mol dm -3 H 2 SO 4 + 2 mol dm -3 VOSO 4 solution. The effects of polarization potential, operating temperature and polarization time on extent of graphite corrosion are investigated by potentiodynamic and potentiostatic techniques. The surface characteristics of graphite electrode before and after corrosion are examined by scanning electron microscopy, atomic force microscopy, and X-ray photoelectron spectroscopy. The results show that the gas begins to evolve on the graphite electrode when the anodic polarization potential is higher than 1.60 V vs saturated calomel electrode at 20 deg. C. The CO 2 evolution on the graphite electrode can lead to intergranular corrosion of the graphite when the polarization potential reaches 1.75 V. In addition, the functional groups of COOH and C=O introduced on the surface of graphite electrode during corrosion can catalyze the formation of CO 2 , therefore, accelerates the corrosion rate of graphite electrode.

  8. 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

  9. 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.

  10. Graphite to Inconel brazing using active filler metal

    International Nuclear Information System (INIS)

    King, J.F.; Baity, F.W.; Walls, J.C.; Hoffman, D.J.

    1989-01-01

    Ion cyclotron resonant frequency (ICRF) antennas are designed to supply large amounts of auxiliary heating power to fusion-grade plasmas in the Toroidal Fusion Test Reactor (TFTR) and Tore Supra fusion energy experiments. A single Faraday shield structure protects a pair of resonant double loops which are designed to launch up to 2 MW of power per loop. The shield consists of two tiers of actively cooled Inconel alloy tubes with the front tier being covered with semicircular graphite tiles. Successful operation of the antenna requires the making of high integrity bonds between the Inconel tubes and graphite tiles by brazing. This paper discusses this process

  11. 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.

  12. Thermal fatigue of refractory metal / graphite composites for fusion applications

    International Nuclear Information System (INIS)

    Smid, I.; Nickel, H.

    1989-01-01

    Reactor grade graphite and molybdenum (TZM) were brazed with different high temperature brazes. The resulting composite tiles had a size of 50 mm x 50 mm with a graphite thickness of 10 mm and a TZM thickness of 5mm. The brazed composites have been tested in electron beam simulation for their thermal fatigue properties. The parameters of these tests were chosen to match NET design specifications for normal operation and 'slow' peak energy deposition. The resulting damages and microstructural changes on the graphites and the brazes are discussed. Additional information is supplied on X-ray diffraction data proving the presence of different phases in the brazes. Finally the influence of a hydrogen plasma on the adaptability of the investigated brazes in fusion devices is discussed. 12 refs., 4 tabs., 4 figs. (Author)

  13. 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.

  14. 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

  15. 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.

  16. A Simulation Model to Evaluate Aircraft Survivability and Target Damage during Offensive Counterair Operations.

    Science.gov (United States)

    1984-03-01

    addressed was a mission o+ two aircraft attacking an area target at an enemy airfi’.fId. The area of opertion: , sias contained within a ten mile radius...sjgb=4.0 if (acft(i,I).gt#1*5)sigb=12.0 c weapon relaibility r6=acft(i,8) c total bombsacft(i,9) nraacft(i,9)/acft(i,5) npacft( i,5) cepacft(i,7

  17. SMART marine goals, targets and management - Is SDG 14 operational or aspirational, is 'Life Below Water' sinking or swimming?

    Science.gov (United States)

    Cormier, Roland; Elliott, Michael

    2017-10-15

    The United Nations Sustainable Development Goals (SDG), adopted in September 2015, are accompanied by targets which have to be met individually and collectively by the signatory states. SDG14 Life Below Water aims to lay the foundation for the integrated and sustainable management of the oceans. However, any environmental management has to be based around targets which are SMART - specific, measurable, achievable, realistic and time bounded - otherwise it is not possible to determine whether management actions are successful and achieve the desired aims. The discussion here shows that many of the targets adopted for SDG14, and especially a detailed analysis of Target 1, are aspirational rather than fully quantified. In order to move towards making the targets operational, we advocate merging the language of environmental management with that used by industry for linking risks to the environment, management performance and ensuing controls. By adopting an approach which uses Key Performance Indicators ('KPIs'), Key Risk Indicators ('KRIs') and Key Control Indicators ('KCIs'), we advocate that a degree of rigour leading to defendable actions can be brought to marine management. Copyright © 2017 Elsevier Ltd. All rights reserved.

  18. Spatially resolved nanostructural transformation in graphite under femtosecond laser irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Marcu, A., E-mail: aurelian.marcu@inflpr.ro [National Institute for Laser, Plasma and Radiation Physics, 077125 Bucharest (Romania); Avotina, L. [Institute of Chemical Physics, University of Latvia, Kronvalda 4, LV 1010 Riga (Latvia); Porosnicu, C. [National Institute for Laser, Plasma and Radiation Physics, 077125 Bucharest (Romania); Marin, A. [Ilie Murgulescu” Institute of Physical Chemistry, 202 Splaiul Independentei 060021, Bucharest (Romania); Grigorescu, C.E.A. [National Institute R& D for Optoelectronics INOE 2000, 077125 Bucharest (Romania); Ursescu, D. [National Institute for Laser, Plasma and Radiation Physics, 077125 Bucharest (Romania); Lungu, M. [National Institute of Materials Physics Atomistilor Str., 105 bis, 077125, Magurele (Romania); Demitri, N. [Hard X-ray Beamline and Structural Biology, Elettra-Sincrotrone Trieste, Strada Statale 14 - km 163,5 in AREA Science Park, 34149 Basovizza TS Italy (Italy); Lungu, C.P. [National Institute for Laser, Plasma and Radiation Physics, 077125 Bucharest (Romania)

    2015-11-15

    Graphical abstract: - Highlights: • Polycrystalline graphite was irradiated with a high power fs (IR) laser. • Presence of a diamond peak was detected by synchrotron XRD. • XPS and Raman showed in-depth sp{sup 3}% increase at tens of nm below the surface. • sp{sup 3}% is increasing with laser power density but it is independent of photon absorption rate. • Graphite crystallite size locally increase at tens of nanometers below the irradiated spots. - Abstract: A polycrystalline graphite target was irradiated using infrared (800 nm) femtosecond (120 fs) laser pulses of different energies. Increase of sp{sup 3} bonds percentage and possible diamond crystal formation were investigated ‘in-depth’ and on the irradiated surfaces. Synchrotron X-ray diffraction pattern have shown the presence of a diamond peak in one of the irradiated zones while X-ray photoelectron spectroscopy investigations have shown an increasing tendency of the sp{sup 3} percent in the low power irradiated areas and similarly ‘in the depth’ of the higher power irradiated zones. Multiple wavelength Micro-Raman investigations have confirmed this trend along with an ‘in-depth’ (but not on the surface) increase of the crystallite size. Based on the wavelength dependent photon absorption into graphite, the observed effects are correlated with high density photon per atom and attributed to the melting and recrystallization processes taking place tens of nanometers below the target surface.

  19. Spatially resolved nanostructural transformation in graphite under femtosecond laser irradiation

    International Nuclear Information System (INIS)

    Marcu, A.; Avotina, L.; Porosnicu, C.; Marin, A.; Grigorescu, C.E.A.; Ursescu, D.; Lungu, M.; Demitri, N.; Lungu, C.P.

    2015-01-01

    Graphical abstract: - Highlights: • Polycrystalline graphite was irradiated with a high power fs (IR) laser. • Presence of a diamond peak was detected by synchrotron XRD. • XPS and Raman showed in-depth sp 3 % increase at tens of nm below the surface. • sp 3 % is increasing with laser power density but it is independent of photon absorption rate. • Graphite crystallite size locally increase at tens of nanometers below the irradiated spots. - Abstract: A polycrystalline graphite target was irradiated using infrared (800 nm) femtosecond (120 fs) laser pulses of different energies. Increase of sp 3 bonds percentage and possible diamond crystal formation were investigated ‘in-depth’ and on the irradiated surfaces. Synchrotron X-ray diffraction pattern have shown the presence of a diamond peak in one of the irradiated zones while X-ray photoelectron spectroscopy investigations have shown an increasing tendency of the sp 3 percent in the low power irradiated areas and similarly ‘in the depth’ of the higher power irradiated zones. Multiple wavelength Micro-Raman investigations have confirmed this trend along with an ‘in-depth’ (but not on the surface) increase of the crystallite size. Based on the wavelength dependent photon absorption into graphite, the observed effects are correlated with high density photon per atom and attributed to the melting and recrystallization processes taking place tens of nanometers below the target surface.

  20. 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

  1. The CNGS target

    CERN Multimedia

    Patrice Loïez

    2005-01-01

    The CERN Neutrinos to Gran Sasso (CNGS) target ‘magazine’ of five target units. Each unit contains a series of 10-cm long graphite rods distributed over a length of 2 m. It is designed to maximize the number of secondary particles produced and hence the number of neutrinos. One unit is used at a time to prevent over heating.

  2. 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.

  3. 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.

  4. Data-Driven Modeling of Target Human Behavior in Military Operations

    Science.gov (United States)

    2014-03-12

    Military Operations Elizabeth Mezzacappa, Ph.D. Gordon Cooke, MEME Gladstone Reid, MSBMS Robert DeMarco, MSBMS Charles Sheridan BA John...stress, and human behavior modeling and simulation issues. GORDON COOKE, MEME , is a Principal Investigator at the TBRL. He was also a Chief

  5. 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.

  6. 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

  7. 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.

  8. 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)

  9. Performance of brazed graphite, carbon-fiber composite, and TZM materials for actively cooled structures: qualification tests

    International Nuclear Information System (INIS)

    Smid, I.; Croessmann, C. D.; Watson, R. D.; Linke, J.; Cardella, A.; Bolt, H.; Reheis, N.; Kny, E.

    1995-01-01

    The divertor of a near-term fusion device has to withstand high heat fluxes, heat shocks, and erosion caused by the plasma. Furthermore, it has to be maintainable through remote techniques. Above all, a good heat removal capability across the interface (low-Z armor/heat sink) plus overall integrity after many operational cycles are needed. To meet all these requirements, an active metal brazing technique is applied to bond graphite and carbon-fiber composite materials to a heat sink consisting of a Mo-41Re coolant tube through a TZM body. Plain brazed graphite and TZM tiles are tested for their fusion-relevant properties. The interfaces appear undamaged after thermal cycling when the melting point of the braze joint is not exceeded and when the graphite armor is > 4 mm thick. High heat flux tests are performed on three actively cooled divertor targets. The braze joints show no sign of failure after exposure to thermal loads ∼ 25 % higher than the design value surface heat flux of 10 MW/m 2 . (author)

  10. Proton-beam window design for a transmutation facility operating with a liquid lead target

    Energy Technology Data Exchange (ETDEWEB)

    Jansen, C.; Lypsch, F.; Lizana, P. [Institute for Safety Research and Reactor Technology, Juelich (Germany)] [and others

    1995-10-01

    The proton beam target of an accelerator-driven transmutation facility can be designed as a vertical liquid lead column. To prevent lead vapor from entering the accelerator vacuum, a proton-beam window has to separate the area above the lead surface from the accelerator tube. Two radiation-cooled design alternatives have been investigated which should withstand a proton beam of 1.6 GeV and 25 mA. Temperature calculations based on energy deposition calculations with the Monte Carlo code HETC, stability analysis and spallation-induced damage calculations have been performed showing the applicability of both designs.

  11. A reverse method for the determination of the radiological inventory of irradiated graphite at reactor scale

    Energy Technology Data Exchange (ETDEWEB)

    Nicaise, Gregory [Institut de Radioprotection et de Surete Nucleaire, Fontenay-aux-roses (France); Poncet, Bernard [EDF-DP2D, Lyon (France)

    2016-11-15

    Irradiated graphite waste will be produced from the decommissioning of the six gas-cooled nuclear reactors operated by Electricite De France (EDF). Determining the radionuclide content of this waste is an important legal commitment for both safety reasons and in order to determine the best suited management strategy. As evidenced by numerous studies nuclear graphite is a very pure material, however, it cannot be considered from an analytical viewpoint as a usual homogeneous material. Because of graphite high purity, radionuclide measurements in irradiated graphite exhibit very high discrepancies especially when corresponding to precursors at trace level. Therefore the assessment of a radionuclide inventory only based on few number of radiochemical measurements leads in most of cases to a gross over or under-estimation that can be detrimental to graphite waste management. A reverse method using an identification calculation-measurement process is proposed in order to assess the radionuclide inventory as precisely as possible.

  12. Inverting the diastereoselectivity of the mukaiyama-michael addition with graphite-based catalysts

    KAUST Repository

    Acocella, Maria Rosaria

    2014-02-07

    Here, we show that graphite-based catalysts, mainly graphite oxide (GO) and exfoliated GO, are effective recyclable catalysts for a relevant stereoselective Mukaiyama-Michael addition, outperforming currently available catalysts. Moreover, the graphite-based catalysts described here invert the diastereoselectivity relative to that observed with known catalysts, with the unprecedented large prevalence of the anti diastereoisomer. This inverted diastereoselectivity is increased when the catalyst concentration is reduced and after catalyst recycling. Density functional theory calculations suggest that the selectivity is determined by two types of supramolecular interactions operating between the catalyst and the substrates at the diastereoselectivity- determining transition state, specifically, the π-stacking of b-nitrostyrene with graphite and the van der Waals interaction between the SiMe3 group of the silyl ether and the graphite. © 2013 American Chemical Society.

  13. The Marine Corps Needs a Targeting, Sensors, and Surveillance Systems Operational Integration and Support Team

    Science.gov (United States)

    2010-03-02

    triggerman is probably still close ; lately all IEDs in the area have been initiated via command-wire. The squad leader sets a cordon, ensures an IED 9...Operational Surveillance System (G-BOSS) with a Class IIIb laser pointer. This class of laser requires users to receive a laser safety class...2) The Keyhole kit of surveillance equipment. Designed to provide “snipers with an increased capability to visually detect the enemy emplacing IEDs

  14. 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.

  15. 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

  16. 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.

  17. '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

  18. 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

  19. Assessment of different mechanisms of C-14 production in irradiated graphite of RBMK-1500 reactors

    International Nuclear Information System (INIS)

    Narkunas, Ernestas; Smaizys, Arturas; Poskas, Povilas; Kilda, Raimondas

    2010-01-01

    Two RBMK-1500 water-cooled graphite-moderated channel-type power reactors at the Ignalina Nuclear Power Plant (INPP) are under decommissioning now. The total mass of irradiated graphite in the cores of both units is more than 3600 tons. The main source of uncertainty in the numerical assessment of graphite activity is the uncertainty of the initial impurities content in graphite. Nitrogen is one of the most important impurities, having a large neutron capture cross-section. This impurity may become the dominant source of C-14 production. RBMK reactors graphite stacks operate in the cooling mixture of helium-nitrogen gases and this may additionally increase the quantity of the nitrogen impurity. In this paper the results of the numerical modelling of graphite activation for the INPP Unit I reactor are presented. In order to evaluate the C-14 activity dependence on the nitrogen impurity content, several cases with different nitrogen content were modelled taking into account initial nitrogen impurity quantities in the graphite matrix and possible nitrogen quantities entrapped in the graphite pores from cooling gases. (orig.)

  20. Grafting polyethylenimine with quinoline derivatives for targeted imaging of intracellular Zn2+ and logic gate operations

    International Nuclear Information System (INIS)

    Pan, Yi; Shi, Yupeng; Chen, Junying; Wong, Chap-Mo; Zhang, Heng; Li, Mei-Jin; Li, Cheuk-Wing; Yi, Changqing

    2016-01-01

    In this study, a highly sensitive and selective fluorescent Zn 2+ probe which exhibited excellent biocompatibility, water solubility, and cell-membrane permeability, was facilely synthesized in a single step by grafting polyethyleneimine (PEI) with quinoline derivatives. The primary amino groups in the branched PEI can increase water solubility and cell permeability of the probe PEIQ, while quinoline derivatives can specifically recognize Zn 2+ and reduce the potential cytotoxicity of PEI. Basing on fluorescence off-on mechanism, PEIQ demonstrated excellent sensing capability towards Zn 2+ in absolute aqueous solution, where a high sensitivity with a detection limit as low as 38.1 nM, and a high selectivity over competing metal ions and potential interfering amino acids, were achieved. Inspired by these results, elementary logic operations (YES, NOT and INHIBIT) have been constructed by employing PEIQ as the gate while Zn 2+ and EDTA as chemical inputs. Together with the low cytotoxicity and good cell-permeability, the practical application of PEIQ in living cell imaging was satisfactorily demonstrated, emphasizing its wide application in fundamental biology research. - Graphical abstract: The fluorescent Zn 2+ probe, PEIQ, is facilely synthesized by grafting PEI with 8-CAAQ, and demonstrated for the pratical applications in Zn 2+ imaging and implementation of molecular logic operations within biological cells. - Highlights: • PEIQ, fluorescent Zn 2+ probe, is synthesized by grafting PEI with quinoline derivatives. • PEIQ exhibits high sensitivity and selectivity in absolute aqueous solution. • PEIQ is biocompatible, water soluble, and cell-membrane permeable. • Elementary logic operations have been demonstrated for PEIQ/Zn 2+ /EDTA system. • The practical application of PEIQ in living cell imaging is demonstrated.

  1. Hunting Leadership Targets in Counterinsurgency and Counterterrorist Operations: Selected Perspectives and Experience

    Science.gov (United States)

    2007-06-01

    DefenceInternet/Defence- News/MilitaryOperations/SpecialForcesSupportGroupFormsInWales. htm. 60. Victor Alonso, “GAL, terrorismo de estado ” (GAL, terrorism of...Guerrilla Movement,” El Financiero , 18 Novem- ber 1994, as translated in FBIS-LAT-94-226, pp. 12-13 c. Orlando Ortíz, Genaro Vázquez, 3rd ed. (Mexico City...Backgrounder No. 660, 5 July 1988 b. Tomás Tenorio Galinda, “Una vieja visión de la guerrilla” c. “The Announced Guerrilla Movement,” El Financiero 97

  2. 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)

  3. HTCAP: a FORTRAN IV program for calculating coated-particle operating temperatures in HFIR target irradiation experiments

    International Nuclear Information System (INIS)

    Kania, M.J.

    1976-05-01

    A description is presented of HTCAP, a computer code that calculates in-reactor operating temperatures of loose coated ThO 2 particles in the HFIR target series of irradiation tests. Three computational models are employed to determine the following: (1) fission heat generation rates, (2) capsule heat transfer analysis, and (3) maximum particle surface temperature within the design of an HT capsule. Maximum particle operating temperatures are calculated at daily intervals during each irradiation cycle. The application of HTCAP to sleeve CP-62 of HT-15 is discussed, and the results are compared with those obtained in an earlier thermal analysis on the same capsule. Agreement is generally within +-5 percent, while decreasing the computational time by more than an order of magnitude. A complete FORTRAN listing and summary of required input data are presented in appendices. Included is a listing of the input data and a tabular output from the thermal analysis of sleeve CP-62 of HT-15

  4. A combined experimental and numerical approach for the control and monitoring of the SPES target during operation at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Ballan, Michele [National Institute of Nuclear Physics – Legnaro National Laboratories (INFN-LNL), Viale dell’Università 2, Legnaro, 35020 Padova (Italy); Department of Physics and Earth Sciences, University of Ferrara, via Saragat 1, 44122 Ferrara (Italy); Manzolaro, Mattia [National Institute of Nuclear Physics – Legnaro National Laboratories (INFN-LNL), Viale dell’Università 2, Legnaro, 35020 Padova (Italy); Department of Industrial Engineering (DII), University of Padova, via Venezia 1, 35131 Padova (Italy); Meneghetti, Giovanni [Department of Industrial Engineering (DII), University of Padova, via Venezia 1, 35131 Padova (Italy); Andrighetto, Alberto [National Institute of Nuclear Physics – Legnaro National Laboratories (INFN-LNL), Viale dell’Università 2, Legnaro, 35020 Padova (Italy); Monetti, Alberto [National Institute of Nuclear Physics – Legnaro National Laboratories (INFN-LNL), Viale dell’Università 2, Legnaro, 35020 Padova (Italy); Department of Industrial Engineering (DII), University of Padova, via Venezia 1, 35131 Padova (Italy); Bisoffi, Giovanni; Prete, Gianfranco [National Institute of Nuclear Physics – Legnaro National Laboratories (INFN-LNL), Viale dell’Università 2, Legnaro, 35020 Padova (Italy)

    2016-06-01

    The SPES project at INFN-LNL aims at the production of neutron-rich Radioactive Ion Beams (RIBs) using the ISOL (Isotope Separation On Line) technique. A 40 MeV 200 μA proton beam will directly impinge a uranium carbide target, generating approximately 10{sup 13} fissions per second. The target system is installed under vacuum inside a water-cooled chamber, and have to maintain high working temperatures, close to 2000 °C. During operation the proton beam provides the heating power required to keep the target at the desired temperature level. As a consequence, its characteristics have to be strictly controlled in order to avoid undesired overheating. According to the original design of the control system, the proton beam can be suddenly interrupted in case of out of range vacuum or cooling water flow levels. With the aim to improve the reliability of the control system a set of temperature sensors has been installed close to the target. Their types and installation positions were defined taking into consideration the detailed information coming from a dedicated thermal–electric model that allowed to investigate the most critical and inaccessible target hot-spots. This work is focused on the definition and experimental validation of the aforementioned numerical model. Its results were used to appropriately install two type C thermocouples, a PT100 thermo-resistance and a residual primary beam current detector. In addition the numerical model will be used for the definition of appropriate thresholds for each installed temperature sensor, since it allows to define a relationship between the locally measured values with the overall calculated temperature field. In case of over temperatures the monitoring system will send warning signals or in case interrupt the proton beam.

  5. A combined experimental and numerical approach for the control and monitoring of the SPES target during operation at high temperature

    Science.gov (United States)

    Ballan, Michele; Manzolaro, Mattia; Meneghetti, Giovanni; Andrighetto, Alberto; Monetti, Alberto; Bisoffi, Giovanni; Prete, Gianfranco

    2016-06-01

    The SPES project at INFN-LNL aims at the production of neutron-rich Radioactive Ion Beams (RIBs) using the ISOL (Isotope Separation On Line) technique. A 40 MeV 200 μA proton beam will directly impinge a uranium carbide target, generating approximately 1013 fissions per second. The target system is installed under vacuum inside a water-cooled chamber, and have to maintain high working temperatures, close to 2000 °C. During operation the proton beam provides the heating power required to keep the target at the desired temperature level. As a consequence, its characteristics have to be strictly controlled in order to avoid undesired overheating. According to the original design of the control system, the proton beam can be suddenly interrupted in case of out of range vacuum or cooling water flow levels. With the aim to improve the reliability of the control system a set of temperature sensors has been installed close to the target. Their types and installation positions were defined taking into consideration the detailed information coming from a dedicated thermal-electric model that allowed to investigate the most critical and inaccessible target hot-spots. This work is focused on the definition and experimental validation of the aforementioned numerical model. Its results were used to appropriately install two type C thermocouples, a PT100 thermo-resistance and a residual primary beam current detector. In addition the numerical model will be used for the definition of appropriate thresholds for each installed temperature sensor, since it allows to define a relationship between the locally measured values with the overall calculated temperature field. In case of over temperatures the monitoring system will send warning signals or in case interrupt the proton beam.

  6. Preliminary design study of a large scale graphite oxidation loop

    International Nuclear Information System (INIS)

    Epel, L.G.; Majeski, S.J.; Schweitzer, D.G.; Sheehan, T.V.

    1979-08-01

    A preliminary design study of a large scale graphite oxidation loop was performed in order to assess feasibility and to estimate capital costs. The nominal design operates at 50 atmospheres helium and 1800 F with a graphite specimen 30 inches long and 10 inches in diameter. It was determined that a simple single walled design was not practical at this time because of a lack of commercially available thick walled high temperature alloys. Two alternative concepts, at reduced operating pressure, were investigated. Both were found to be readily fabricable to operate at 1800 F and capital cost estimates for these are included. A design concept, which is outside the scope of this study, was briefly considered

  7. 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

  8. 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

  9. 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.

  10. 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

  11. Design and performance of a clean and economic power supply targeting energy independence and operative security

    Energy Technology Data Exchange (ETDEWEB)

    Mushtaha, Mohammed

    2012-07-01

    Stand alone wind-diesel power systems address the major issues for the supply of remotely located consumers or the provision of enhanced energy independence. The specific configuration considered here comprises a prior wind turbine driven induction generator and a supplementary synchronous generator driven by a combustion engine, the latter caring for frequency control and reactive power balance. An accumulator-bank and a hydrogen storage path consisting of an electrolyzer and a hydrogen storage tank are added as short-term and long-term storage devices - respectively - in order to adjust volatile wind generation and varying load demand. The hydrogen re-conversion to electricity is achieved by design of the drive of the synchronous generator as a dual-fuel engine, fed with stored hydrogen primarily, and with diesel oil in case of the rare event of hydrogen storage emptiness. An elementary waste heat recovery from engine cooling water and from exhaust gases makes the system even more energy efficient and environment friendly. For proper sizing of the particular system components under minimization of the life cycle cost - which are composed of the initial installation cost, the replacement cost, the operation and maintenance cost as well as the remaining fuel cost - the meta-heuristic computational method of Particle Swarm Optimization (PSO) was applied. The crucial task of intelligently managing the energy storage strategy depending on their actual states of charge and the current excess power under a given load profile was solved by a fuzzy system and a subsequent time scheduler, the latter providing definite set-point steps for definite periods of time to the electrolyzer in order to achieve smooth and conservative operation. The concrete design and performance of the proposed system structure is shown with the two relevant application examples of independent electricity supply of the main hospital in Gaza-Strip, Palestine, and a remotely located dairy farm. For

  12. Thickness gauge for the measurement of the density of graphite

    International Nuclear Information System (INIS)

    Leveque, P.; Gasnier, M.; Hours, R.; Jouquet, G.; Rappeneau, J.; Tanguy, J.C.

    1961-01-01

    A thickness gauge was built, based on absorption of Bremsstrahlung generated in a Be target by a ( 90 Sr + 90 Y) β- source. This allows rapid and precise estimation (95 per cent probable error = 0.7 per cent) of the densities in slabs of graphite having a constant thickness of 25 ± 0.05 mm, the diameter of the beam being about 1 cm. Results obtained in this way are presented. (author) [fr

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-05-01

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

  15. 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

  16. 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.

  17. 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)

  18. 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

  19. 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)

  20. Graphite suspension in carbon dioxide; Suspension de graphite dans le gaz carbonique

    Energy Technology Data Exchange (ETDEWEB)

    Roche, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Moussez, C; Rouvillois, X; Brevet, R [Societe Nationale d' Etude et de Construction de Moteurs d' Aviation (SNECMA), 75 - Paris (France)

    1965-07-01

    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{sup 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{sup 2}/g (graphite particles about 1 {mu}), the powder surface area reaches an asymptotic value of 300 m{sup 2}/g (all the particles less than 0.3 {mu}). Moisture effect on flow stability, flow distribution between two parallel channels, pressure drop in straight tubes, recompression ratio in diffusers were also investigated. (author) [French] Depuis 1963 la Division Atomique de la SNECMA conduit, dans le cadre d'un contrat avec le Commissariat A l'Energie Atomique, l'etude experimentale d'une suspension de fines particules de graphite dans le gaz carbonique. L'objectif principal est d'obtenir des informations d'ordre mecanique et technologique sur la mise en oeuvre de l'ecoulement de ce fluide diphase. Le circuit experimental comprend principalement: un

  1. Synthesis and characterization of SiC based composite materials for immobilizing radioactive graphite

    Science.gov (United States)

    Wang, Qing; Teng, Yuancheng; Wu, Lang; Zhang, Kuibao; Zhao, Xiaofeng; Hu, Zhuang

    2018-06-01

    In order to immobilize high-level radioactive graphite, silicon carbide based composite materials{ (1-x) SiC· x MgAl2O4 (0.1 ≤ x≤0.4) } were fabricated by solid-state reaction at 1370 °C for 2 h in vacuum. Residual graphite and precipitated corundum were observed in the as-synthesized product, which attributed to the interface reaction of element silicon and magnesium compounds. To further understand the reasons for the presence of graphite and corundum, the effects of mole ratio of Si/C, MgAl2O4 content and non-stoichiometry of MgAl2O4 on the synthesis were investigated. To immobilize graphite better, residual graphite should be eliminated. The target product was obtained when the mole ratio of Si/C was 1.3:1, MgAl2O4 content was x = 0.2, and the mole ratio of Al to Mg in non-stoichiometric MgAl2O4 was 1.7:1. In addition, the interface reaction between magnesium compounds and silicon not graphite was displayed by conducting a series of comparative experiments. The key factor for the occurrence of interface reaction is that oxygen atom is transferred from magnesium compound to SiO gas. Infrared and Raman spectrum revealed the increased disorders of graphite after being synthesized.

  2. Oxidation Behavior of IG-11, IG-110 and IG-430 Graphites in Air Flow

    International Nuclear Information System (INIS)

    Hong, Jin Ki; Chi, Se Hwan

    2006-01-01

    In high temperature gas-cooled reactor (HTGR), graphite is used as a moderator and a reflector as well as a major structural component. During operation or in the event of an accident, subsequent graphite oxidation due to the graphite out-gassing or heat exchanger tube leakage results in changes in the physical and mechanical properties of the components. For this reason, a lot of studies on oxidation have long been performed to understand the high temperature oxidation behavior and to find a proper countermeasure over the expected operating range. In this study, the oxidation rates of IG-11, IG-110 and IG-430 nuclear graphites were determined at high temperature and evaluated in view of the grades and the oxidation mechanisms at different temperature range

  3. Graphite stack corrosion of BUGEY-1 reactor (synthesis)

    International Nuclear Information System (INIS)

    Petit, A.; Brie, M.

    1996-01-01

    The definitive shutdown date for the BUGEY-1 reactor was May 27th, 1994, after 12.18 full power equivalent years and this document briefly describes some of the feedback of experience from operation of this reactor. The radiolytic corrosion of graphite stack is the major problem for BUGEY-1 reactor, despite the inhibition of the reaction by small quantities of CH 4 added to the coolant gas. The mechanical behaviour of the pile is predicted using the ''INCA'' code (stress calculation), which uses the results of graphite weight loss variation determined using the ''USURE'' code. The weight loss of graphite is determined by annually taking core samples from the channel walls. The results of the last test programme undertaken after the definitive shutdown of BUGEY-1 have enabled an experimental graph to be established showing the evolution of the compression resistance (perpendicular and parallel direction to the extrusion axis) as a function of the weight loss. The numerous analyses, made on the samples carried out in the most sensitive regions, have allowed to verify that no brutal degradation of the mechanical properties of graphite happens for the high value of weight loss up to 40% (maximum weight loss reached locally). (author). 10 refs, 3 figs, 4 tabs

  4. IAEA international database on irradiated nuclear graphite properties

    International Nuclear Information System (INIS)

    Burchell, T.D.; Clark, R.E.H.; Stephens, J.A.; Eto, M.; Haag, G.; Hacker, P.; Neighbour, G.B.; Janev, R.K.; Wickham, A.J.

    2000-02-01

    This report describes an IAEA database containing data on the properties of irradiated nuclear graphites. Development and implementation of the graphite database followed initial discussions at an IAEA Specialists' Meeting held in September 1995. The design of the database is based upon developments at the University of Bath (United Kingdom), work which the UK Health and Safety Executive initially supported. The database content and data management policies were determined during two IAEA Consultants' Meetings of nuclear reactor graphite specialists held in 1998 and 1999. The graphite data are relevant to the construction and safety case developments required for new and existing HTR nuclear power plants, and to the development of safety cases for continued operation of existing plants. The database design provides a flexible structure for data archiving and retrieval and employs Microsoft Access 97. An instruction manual is provided within this document for new users, including installation instructions for the database on personal computers running Windows 95/NT 4.0 or higher versions. The data management policies and associated responsibilities are contained in the database Working Arrangement which is included as an Appendix to this report. (author)

  5. Project accent: graphite irradiated creep in a materials test reactor

    International Nuclear Information System (INIS)

    Brooking, M.

    2014-01-01

    Atkins manages a pioneering programme of irradiation experiments for EDF Energy. One of these projects is Project ACCENT, designed to obtain evidence of a beneficial physical property of the graphite, which may extend the life of the Advanced Gas-cooled Reactors (AGRs). The project team combines the in-house experience of EDF Energy with two supplier organisations (providing the material test reactors and testing facilities) and supporting consultancies (Atkins and an independent technical expert). This paper describes: - Brief summary of the Project; - Discussion of the challenges faced by the Project; and - Conclusion elaborating on the aims of the Project. These challenging experiments use bespoke technology and both un-irradiated (virgin) and irradiated AGR graphite. The results will help to better understand graphite irradiation-induced creep (or stress modified dimensional change) properties and therefore more accurately determine lifetime and safe operating envelopes of the AGRs. The first round of irradiation has been completed, with a second round about to commence. This is a key step to realising the full lifetime ambition for AGRs, demonstrating the relaxation of stresses within the graphite bricks. (authors)

  6. 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.

  7. 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

  8. Why Won't You Drop, Damn YouZ? An Examination of the Targeting Process in Operation Enduring Freedom and its Implications

    National Research Council Canada - National Science Library

    Kindley, David

    2004-01-01

    ...?. This paper will examine the decision making process for rapid targeting in OEF from the strike fighter pilot through the watch officer in the Combined Air Operations Center at Prince Sultan Air Base...

  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. 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.

  12. 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)

  13. Numerical and Experimental Investigation on the Spray Coating Process Using a Pneumatic Atomizer: Influences of Operating Conditions and Target Geometries

    Directory of Open Access Journals (Sweden)

    Qiaoyan Ye

    2017-01-01

    Full Text Available This paper presents a numerical simulation of the spray painting process using a pneumatic atomizer with the help of a computational fluid dynamics code. The droplet characteristics that are necessary for the droplet trajectory calculation were experimentally investigated using different shaping air flow rates. It was found that the droplet size distribution depends on both the atomizing and the shaping air flow rate. An injection model for creating the initial droplet conditions is necessary for the spray painting simulation. An approach for creating these initial conditions has been proposed, which takes different operating conditions into account and is suitable for practical applications of spray coating simulation using spray guns. Further, tests on complicated targets and complex alignments of the atomizer have been carried out to verify this numerical approach. The results confirm the applicability and reliability of the chosen method for the painting process.

  14. 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)

  15. 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

  16. Observation of non-uniform erosion and deposition phenomena on graphite after plasma exposure

    International Nuclear Information System (INIS)

    Hildebrandt, D.; Grote, H.; Schneider, W.; Wienhold, P.; Seggern, J. von

    1999-01-01

    The modifications of fine grain isotropic graphite surfaces after plasma exposure have been investigated using surface analysis techniques with high spatial resolution in area and depth. The samples are graphite target tiles of ASDEX-upgrade and coated graphite collector samples exposed for special erosion/deposition experiments in the divertor plasma of ASDEX-upgrade or in the scrape-off plasma of TEXTOR-94. In addition, a graphite sample was exposed to a low temperature, clean deuterium plasma to study the modifications of the surface morphology during plasma exposure. The results give clear indications of non-uniform erosion and deposition processes. The change of the surface morphology during these processes is discussed. (orig.)

  17. Optimization of Sealed Tube Graphitization Method for Environmental C-14 Studies Using MICADAS

    Energy Technology Data Exchange (ETDEWEB)

    Rinyu, Laszlo, E-mail: rinyu@atomki.hu [Hertelendi Laboratory of Environmental Studies, Institute of Nuclear Research of the Hungarian Academy of Sciences (ATOMKI), H-4026 Debrecen (Hungary); Isotoptech Zrt., H-4025 Debrecen (Hungary); Molnar, Mihaly [Hertelendi Laboratory of Environmental Studies, Institute of Nuclear Research of the Hungarian Academy of Sciences (ATOMKI), H-4026 Debrecen (Hungary); Ion Beam Physics, ETH Zuerich, CH-8093 Zuerich (Switzerland); Major, Istvan [Hertelendi Laboratory of Environmental Studies, Institute of Nuclear Research of the Hungarian Academy of Sciences (ATOMKI), H-4026 Debrecen (Hungary); Nagy, Tamas; Veres, Mihaly [Isotoptech Zrt., H-4025 Debrecen (Hungary); Kimak, Adam [University of Debrecen, H-4032 Debrecen (Hungary); Wacker, Lukas; Synal, Hans-Arno [Ion Beam Physics, ETH Zuerich, CH-8093 Zuerich (Switzerland)

    2013-01-15

    The original sealed tube zinc reduction graphitization process was first developed for rapid low-precision measurements of biomedical tracer samples and later also applied for high precision measurements of not too old samples. In this study we tested the MICADAS (mini radiocarbon dating system [1]) radiocarbon measurements of targets prepared by sealed tube graphitization process. We found the optimal iron catalyst and reagents (TiH{sub 2} and Zn) amount whereby we can reach a relatively low background level, and minimized the overall {delta}{sup 13}C fractionation during the graphitization. Repeated measurements of normalization standards and real samples with known {sup 14}C activities were very well reproduced. Finally, we demonstrated the applicability of the sealed tube graphitization on real environmental samples covering a wide range of {sup 14}C concentrations.

  18. Molecular dynamics simulation of chemical sputtering of hydrogen atom on layer structured graphite

    International Nuclear Information System (INIS)

    Ito, A.; Wang, Y.; Irle, S.; Morokuma, K.; Nakamura, H.

    2008-10-01

    Chemical sputtering of hydrogen atom on graphite was simulated using molecular dynamics. Especially, the layer structure of the graphite was maintained by interlayer intermolecular interaction. Three kinds of graphite surfaces, flat (0 0 0 1) surface, armchair (1 1 2-bar 0) surface and zigzag (1 0 1-bar 0) surface, are dealt with as targets of hydrogen atom bombardment. In the case of the flat surface, graphene layers were peeled off one by one and yielded molecules had chain structures. On the other hand, C 2 H 2 and H 2 are dominant yielded molecules on the armchair and zigzag surfaces, respectively. In addition, the interaction of a single hydrogen isotope on a single graphene is investigated. Adsorption, reflection and penetration rates are obtained as functions of incident energy and explain hydrogen retention on layered graphite. (author)

  19. Assessments of the stresses and deformations in an RBMK graphite moderator brick

    International Nuclear Information System (INIS)

    Jones, C.J.; Davies, M.A.; Marsden, B.J.; Bougaenko, S.E.; Baldin, V.D.; Demintievski, V.N.; Rodtchenkov, B.S.; Sinitsyn, E.N.

    1996-01-01

    The RBMK reactors, designed by RDIPE (Moscow), are graphite moderated and cooled by light water. Graphite dimensions and thermo-mechanical properties change significantly in a complex manner during reactor life due to fast neutron damage and these changes have implications on the safe operation of all graphite moderated reactors. A joint programme of work is being carried out between AEA Technology (UK) and RDIPE (Russia) to assess the life of the RBMK graphite stack under normal operating conditions. The programme has included the modelling of graphite dimensional changes due to irradiation through reactor life and the assessment of the implications of these changes on the stresses and deformations in the graphite stack. Calculations have been carried out to assess the deformations of a moderator brick over a period from start of life up to 30 years of operation. The assessment have also included an analysis of the stresses in the bricks so that the time to brick failure could be determined. This paper describes the RBMK core design, the data and assessment methodology used in the analysis of the RBMK core and presents some results from analyses of the Leningrad Unit 1 RBMK reactor. (author). 2 refs, 8 figs

  20. 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)

  1. 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.

  2. 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

  3. 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.

  4. 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

  5. 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

  6. 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.

  7. 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.

  8. 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

  9. Estimation of Graphite Dust Production in ITER TBM

    International Nuclear Information System (INIS)

    Kang, Ji Ho; Kim, Eung Seon

    2013-01-01

    This scheme uses simple equations and the calculation time is much less than others. However, the contact equation requires a specially tuned material properties and instability of system matrix were reported. Second, only a couple of pebbles were modeled using FEM(Finite Element Method) and appropriate boundary and loading conditions are imposed. This scheme gives a detailed information of stress distribution of the pebbles and the stability of calculation is well established. However, the calculation cost is fairly high and only a few pebble can be analyzed in detail at a time with specifically assigned contact conditions. In this study, a prediction model of graphite dust production in ITER(International Thermonuclear Experimental Reactor) TBM(Test Blanket Module) using FEM was introduced and the amount of dust production for an operation cycle was estimated. In this study, graphite dust generation in the reflector zone of ITER TBM was estimated using FE analysis. A unit-cell model was defined to simulate normal contact forces and slip distances on contact points between the center pebble and the surrounding pebbles. The dust production was calculated using Archard equation. The simulation was repeated with different friction coefficient of graphite material to investigate the effect of friction on the dust production. The calculation result showed that the amount of dust production was 2.22∼3.67e-4 g/m 3 which was almost linearly proportional to the friction coefficient of graphite material. The amount of graphite dust production was considered too much small for a dust explosion

  10. Modelling of pulsed electron beam induced graphite ablation: Sublimation versus melting

    Science.gov (United States)

    Ali, Muddassir; Henda, Redhouane

    2017-12-01

    Pulsed electron beam ablation (PEBA) has recently emerged as a very promising technique for the deposition of thin films with superior properties. Interaction of the pulsed electron beam with the target material is a complex process, which consists of heating, phase transition, and erosion of a small portion from the target surface. Ablation can be significantly affected by the nature of thermal phenomena taking place at the target surface, with subsequent bearing on the properties, stoichiometry and structure of deposited thin films. A two stage, one-dimensional heat conduction model is presented to describe two different thermal phenomena accounting for interaction of a graphite target with a polyenergetic electron beam. In the first instance, the thermal phenomena are comprised of heating, melting and vaporization of the target surface, while in the second instance the thermal phenomena are described in terms of heating and sublimation of the graphite surface. In this work, the electron beam delivers intense electron pulses of ∼100 ns with energies up to 16 keV and an electric current of ∼400 A to a graphite target. The temperature distribution, surface recession velocity, ablated mass per unit area, and ablation depth for the graphite target are numerically simulated by the finite element method for each case. Based on calculation findings and available experimental data, ablation appears to occur mainly in the regime of melting and vaporization from the surface.

  11. 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) ...

  12. Antiproton Target

    CERN Multimedia

    1980-01-01

    Antiproton target used for the AA (antiproton accumulator). The first type of antiproton production target used from 1980 to 1982 comprised a rod of copper 3mm diameter and 120mm long embedded in a graphite cylinder that was itself pressed into a finned aluminium container. This assembly was air-cooled and it was used in conjunction with the Van der Meer magnetic horn. In 1983 Fermilab provided us with lithium lenses to replace the horn with a view to increasing the antiproton yield by about 30%. These lenses needed a much shorter target made of heavy metal - iridium was chosen for this purpose. The 50 mm iridium rod was housed in an extension to the original finned target container so that it could be brought very close to the entrance to the lithium lens. Picture 1 shows this target assembly and Picture 2 shows it mounted together with the lithium lens. These target containers had a short lifetime due to a combination of beam heating and radiation damage. This led to the design of the water-cooled target in...

  13. 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...

  14. Fine-Grained Targets for Laser Synthesis of Carbon Nanotubes

    Science.gov (United States)

    Smith, Michael W. (Inventor); Park, Cheol (Inventor)

    2017-01-01

    A mechanically robust, binder-free, inexpensive target for laser synthesis of carbon nanotubes and a method for making same, comprising the steps of mixing prismatic edge natural flake graphite with a metal powder catalyst and pressing the graphite and metal powder mixture into a mold having a desired target shape.

  15. Irradiation damage in graphite due to fast neutrons in fission and fusion systems

    International Nuclear Information System (INIS)

    2000-09-01

    Gas cooled reactors have been in operation for the production of electricity for over forty years, encompassing a total of 56 units operated in seven countries. The predominant experience has been with carbon dioxide cooled reactors (52 units), with the majority operated in the United Kingdom. In addition, four prototype helium cooled power plants were operated in the United States and Germany. The United Kingdom has no plans for further construction of carbon dioxide units, and the last helium cooled unit was shutdown in 1990. However, there has been an increasing interest in modular helium cooled reactors during the 1990s as a possible future nuclear option. Graphite is a primary material for the construction of gas cooled reactor cores, serving as a low absorption neutron moderator and providing a high temperature, high strength structure. Commercial gas cooled reactor cores (both carbon dioxide cooled and helium cooled) utilise large quantities of graphite. The structural behaviour of graphite (strength, dimensional stability, susceptibility to cracking, etc.) is a complex function of the source material, manufacturing process, chemical environment, and temperature and irradiation history. A large body of data on graphite structural performance has accumulated from operation of commercial gas cooled reactors, beginning in the 1950s and continuing to the present. The IAEA is supporting a project to collect graphite data and archive it in a retrievable form as an International Database on Irradiated Nuclear Graphite Properties, with limited general access and more detailed access by participating Member States. Because of the large size of the database, the complexity of the phenomena and the number of variables involved, a general understanding of graphite behaviour is essential to the understanding and use of the data

  16. 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.

  17. Evaluation of high temperature brazes for graphite first wall protection elements

    International Nuclear Information System (INIS)

    Smid, I.; Koizlik, K.; Linke, J.; Nickel, H.; Wallura, E.; Kny, E.

    1989-01-01

    Four different high temperature brazed with melting points from 800 to 1865degC have been used to braze a commercial reactor grade graphite to TZM substrates. Those brazes were Zr, 90Ni 10Ti, 99Cu 10Ti and 70Ag 27Cu 3Ti (wt %). The resulting composite tiles of 80 x 80 mm 2 with a graphite thickness of 10 mm brazed on a 8 mm TZM substrate have been tested in electron beam experiments for their thermal fatigue properties. The parameters of the electron beam testing were chosen to match NET design specificatios for normal operation and 'slow' peak energy deposition. The resulting damages and microstructural changes on the graphite and the brazes are discussed. Additional information is supplied on tensile test and thermal conductivity data of brazed composites. These measurements confirm that thermal contact between TZM-substrate and graphite is improved by brazing. (author). 6 refs.; 5 figs.; 2 tabs

  18. Evaluation of high temperature brazes for graphite first wall protection elements

    International Nuclear Information System (INIS)

    Smid, I.; Koizlik, K.; Linke, J.; Nickel, H.; Wallura, E.; Kny, E.

    1995-01-01

    Four different high temperature brazes with melting points from 800 to 1865 degree C have been used to braze a commercial reactor grade graphite to TZM substrates. Those brazes were Zr, 90Ni 10Ti, 90Cu 10Ti and 70Ag 27Cu 3Ti (wt %). The resulting composite tiles of 80 x 80 mm 2 with a graphite thickness of 10 mm brazed on a 3 mm TZM substrate have been tested in electron beam experiments for their thermal fatigue properties. The parameters of the electron beam testing were chosen to match NET design specifications for normal operation and 'slow' peak energy deposition. The resulting damages and microstructural changes on the graphite and the brazes are discussed. Additional information is supplied on tensile test and thermal conductivity data of brazed composites. These measurements confirm that thermal contact between TZM-substrate and graphite is improved by brazing. (author)

  19. Depleted Hydrocarbon Reservoirs Present a Safe and Practical Burial Solution for Graphite Waste

    International Nuclear Information System (INIS)

    Rahmani, L.

    2016-01-01

    A solution for graphite waste is proposed that combines reliance on thick impermeable host rock that is needed to confine the long-life radioactivity content of most irradiated graphite with low capitalistic and operational unit volume costs that are required to render this bulky waste form manageable. The solution, uniquely applicable to irradiated graphite due to its low dose rates, moderate mechanical strength and light density, consists in three steps: first, graphite is fine-crushed under water; second, it is made in an aqueous suspension; third, the suspension is injected into a deep, disused hydrocarbon reservoir. Each of these steps only involves well mastered techniques. Regulatory changes that may allow this solution to be added to the gamut of available waste routes, geochemical issues, availability of depleted reservoirs and cost projections are presented. (author)

  20. Measurements of impurity migration in graphite at high temperatures using a proton microprobe

    International Nuclear Information System (INIS)

    Shroy, R.E.; Soo, P.; Sastre, C.A.; Schweiter, D.G.; Kraner, H.W.; Jones, K.W.

    1978-01-01

    The migration of fission products and other impurities through the graphite core of a High Temperature Gas Cooled Reactor is of prime importance in studies of reactor safety. Work in this area is being carried out in which graphite specimens are heated to temperatures up to 3800 0 C to induce migration of trace elements whose local concentrations are then measured with a proton microprobe. This instrument is a powerful device for such work because of its ability to determine concentrations at a part per million (ppm) level in a circular area as small as 10 μm while operating in an air environment. Studies show that Si, Ca, Cl, and Fe impurities in graphite migrate from hotter to cooler regions. Also Si, S, Cl, Ca, Fe, Mn, and Cr are observed to escape from the graphite and be deposited on cooler surfaces

  1. 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.

  2. 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)

  3. 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...

  4. 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)

  5. 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.

  6. 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.

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

    International Nuclear Information System (INIS)

    Le-Guillou, Mael

    2014-01-01

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

  8. Examination of Performance Management Targets in Case of an International Corporation’s Eastern Hungarian Operating Unit

    Directory of Open Access Journals (Sweden)

    Nemeth Zoltan

    2016-06-01

    Full Text Available Nowadays, due to the social and economic changes, the human is already the main resource which determines the successfulness of corporations. The expertise in managing the human resources is a key factor of competitiveness therefore it has become a strategic question. Primary aim of the treatise is to outline the most important principles, targets and peculiarities of performance management system, which is a central element of human resource management, as well as its interconnectedness with the individual performance evaluation and human resource development through the mirror of one of today’s biggest Hungarian economic corporations. We aimed to explore, by means of document analysis as well as methods of case study and interview, how the targets of performance management are used for increasing the economic productivity, preserving the competitiveness, retaining and enhancing the satisfaction of workers. It can be stated that the PM system was continuously improved in case of the organization examined because the corporate goals and the expected performances were adjusted to the changing market demands over time but, at the same time the goal of differentiation has remained unchanged. Each worker evaluates his/her performance individually, the base of which is the extent of contribution to the corporate results. Based on the results, it can overall be stated that such a well-functioning performance management system is applied in the operating unit of the corporation examined which is strongly supported by the organizational culture and which contributes to the joint fulfilment of personal as well as organizational goals.

  9. 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

  10. 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

  11. 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)

  12. 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

  13. On the development of standardization methods for measuring the degree of graphitization of industrial materials, correlation of scientific characterization with the industrially relevant secondary properties of graphitic material

    International Nuclear Information System (INIS)

    Fitzer, E.; Koechling, K.H.

    1984-01-01

    The completed research roject comprised the development of standardization methods for measuring the degree of graphitization of industrial materials, the testing of characterization methods with regard to significance, the correlation of scientific characterization with industrially relevant material properties and the resultant modification of industrial processes for the manufacture of coke raw materials and graphitic materials. Also targeted within the scope of this project were the grouping of comparable characterization methods and the drawing up of unequivocal terminology for scientific and industrial-commercial usage, with both aims based on intensified national and international teamwork. (orig./WL) [de

  14. Graphitization of diamond with a metallic coating on ferritic matrix; Grafitizacao do diamante com revestimento metalico em matriz ferritica

    Energy Technology Data Exchange (ETDEWEB)

    Cabral, Stenio Cavalier; Oliveira, Hellen Cristine Prata de; Filgueira, Marcello, E-mail: stenio@uenf.b [Universidade Estadual do Norte Fluminense (PPGECM/CCT/UENF), Campos dos Goytacazes, RJ (Brazil). Centro de Ciencias e Tecnologia. Programa de Pos Graduacao em Engenharia e Ciencia dos Materiais

    2010-07-01

    Iron is a strong catalyst of graphitization of diamonds. This graphitization occurs mainly during the processing of composites - conventional sintering or hot pressing, and during cutting operations. Aiming to avoid or minimize this deleterious effect, there is increasing use of diamond coated with metallic materials in the production of diamond tools processed via powder metallurgy. This work studies the influence of Fe on diamond graphitization diamond-coated Ti after mixing of Fe-diamonds, hot pressing parameters were performed with 3 minutes/35MPa/900 deg C - this is the condition of pressing hot used in industry for production of diamond tools. Microstructural features were observed by SEM, diffusion of Fe in diamond was studied by EDS. Graphitization was analyzed by X-ray diffraction and Raman spectroscopy. It was found that Fe not activate graphitization on the diamond under the conditions of hot pressing. (author)

  15. EDF - 2015 full-year results: all targets reached, Strong operating performance in adverse market conditions, 2018 ambition reiterated

    International Nuclear Information System (INIS)

    2016-01-01

    A key player in energy transition, the EDF Group is an integrated electricity company, active in all areas of the business: generation, transmission, distribution, energy supply and trading, energy services. A global leader in low-carbon energies, the Group has developed a diversified generation mix based on nuclear power, hydropower, new renewable energies and thermal energy. The Group is involved in supplying energy and services to approximately 37.6 million customers, of which 27.8 million in France. The Group generated consolidated sales of Euro 75 billion in 2015, of which 47.2% outside of France. EDF is listed on the Paris Stock exchange. EDF achieved all its targets in 2015. The year was marked by strong operational performance, reflecting the significant efforts the teams made. With the end of the regulated Yellow and Green Tariffs, most clients turned to EDF. Nuclear output reached its highest level, since 2011 in France, and since 2005 in the United Kingdom. EDF is also continuing its significant development in renewable energy, with an additional 1 GW of net installed capacity. The transformation of EDF Group is essential in the unfavourable market conditions. EDF has embarked on this transformation, and is accelerating innovation to serve the energy transition

  16. 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)

  17. 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

  18. Source Term Analysis of the Irradiated Graphite in the Core of HTR-10

    Directory of Open Access Journals (Sweden)

    Xuegang Liu

    2017-01-01

    Full Text Available The high temperature gas-cooled reactor (HTGR has potential utilization due to its featured characteristics such as inherent safety and wide diversity of utilization. One distinct difference between HTGR and traditional pressurized water reactor (PWR is the large inventory of graphite in the core acting as reflector, moderator, or structure materials. Some radionuclides will be generated in graphite during the period of irradiation, which play significant roles in reactor safety, environmental release, waste disposal, and so forth. Based on the actual operation of the 10 MW pebble bed high temperature gas-cooled reactor (HTR-10 in Tsinghua University, China, an experimental study on source term analysis of the irradiated graphite has been done. An irradiated graphite sphere was randomly collected from the core of HTR-10 as sample in this study. This paper focuses on the analytical procedure and the establishment of the analytical methodology, including the sample collection, graphite sample preparation, and analytical parameters. The results reveal that the Co-60, Cs-137, Eu-152, and Eu-154 are the major γ contributors, while H-3 and C-14 are the dominating β emitting nuclides in postirradiation graphite material of HTR-10. The distribution profiles of the above four nuclides are also presented.

  19. Non-activated high surface area expanded graphite oxide for supercapacitors

    Energy Technology Data Exchange (ETDEWEB)

    Vermisoglou, E.C.; Giannakopoulou, T.; Romanos, G.E.; Boukos, N.; Giannouri, M. [Institute of Nanoscience and Nanotechnology “Demokritos”, 153 43 Ag. Paraskevi, Attikis (Greece); Lei, C.; Lekakou, C. [Division of Mechanical, Medical, and Aerospace Engineering, Faculty of Engineering and Physical Sciences, University of Surrey, Guildford GU2 7XH (United Kingdom); Trapalis, C., E-mail: c.trapalis@inn.demokritos.gr [Institute of Nanoscience and Nanotechnology “Demokritos”, 153 43 Ag. Paraskevi, Attikis (Greece)

    2015-12-15

    Graphical abstract: - Highlights: • One-step exfoliation and reduction of graphite oxide via microwave irradiation. • Effect of pristine graphite (type, flake size) on the microwave expanded material. • Effect of pretreatment and oxidation cycles on the produced expanded material. • Expanded graphene materials with high BET surface areas (940 m{sup 2}/g–2490 m{sup 2}/g). • Non-activated graphene based materials suitable for supercapacitors. - Abstract: Microwave irradiation of graphite oxide constitutes a facile route toward production of reduced graphene oxide, since during this treatment both exfoliation and reduction of graphite oxide occurs. In this work, the effect of pristine graphite (type, size of flakes), pretreatment and oxidation cycles on the finally produced expanded material was examined. All the types of graphite that were tested afforded materials with high BET surface areas ranging from 940 m{sup 2}/g to 2490 m{sup 2}/g, without intervening an activation stage at elevated temperature. SEM and TEM images displayed exfoliated structures, where the flakes were significantly detached and curved. The quality of the reduced graphene oxide sheets was evidenced both by X-ray photoelectron spectroscopy and Raman spectroscopy. The electrode material capacitance was determined via electrochemical impedance spectroscopy and cyclic voltammetry. The materials with PEDOT binder had better performance (∼97 F/g) at low operation rates while those with PVDF binder performed better (∼20 F/g) at higher rates, opening up perspectives for their application in supercapacitors.

  20. 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)

  1. 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

  2. Low-temperature H2-4He and H2-3He targets for operation on an electron beam

    International Nuclear Information System (INIS)

    Gol'dshtejn, V.A.; Lubyanyj, V.V.

    1981-01-01

    Structures and basic characteristics of H 2 - 4 He and H 2 - 3 He low temperature targets are given. Technique of 3 He target filling is described. Initial target cooling up to liquid 4 He temperature and its filling up take near approximately 1 h, at that 4 He flow rate equals 15 l. Repeated filling up of 4 He takes 20 min, and target filling up with 3 He - 10-15 min. Good thermal insulation of a cryostat and targets permits the 4 He target to be operated with an electron beam of a mean current of up to 0.5 μA without filling up 4 He for 70 h. At that flow rate of liquid 4 He amounts to 0.2 l/h, and liquid hydrogen - 0.04 l/h. It is concluded that H 2 - 4 He and H 2 - 3 He targets are reliable and simple in operation and permit to work with accelerated particle beams of intensity corresponding to power release >= 0.5 W without corrections for density change [ru

  3. 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

  4. 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.

  5. 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.

  6. 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

  7. Product Evaluation Task Force Phase Two report for CAGR graphite

    International Nuclear Information System (INIS)

    Francis, A.J.; Davies, A.

    1991-01-01

    It has been proposed that all Intermediate Level Wastes arising at Sellafield should be encapsulated prior to ultimate disposal. The Product Evaluation Task Force (PETF) was set up to investigate possible encapsulants and to produce an adequate data base to justify the preferred matrices. This report details the work carried out under Phase 2 of the Product Evaluation Task Force programme, on CAGR graphite. Three possible types of encapsulants for CAGR graphites:-Inorganic cements, Polymer cements and Polymers are evaluated using the Kepner Tregoe decision analysis technique. This technique provides a methodology for scoring and ranking alternative options and evaluating any risks associated with an option. The analysis shows that for all four stages of waste management operations ie Storage, Transport, handling and emplacement, Disposal and Process, cement matrices are considerably superior to other potential matrices. A matrix, consisting of three parts Blast Furnace Slag (BFS) to one part Ordinary Portland Cement (OPC) is recommended as the preferred matrix for Phase 3 studies on CAGR graphite. (author)

  8. Graphite core stability during 'care and maintenance' and 'safe storage'

    International Nuclear Information System (INIS)

    Wickham, A.J.; Marsden, B.J.; Sellers, R.M.; Pilkington, N.J.

    1998-01-01

    The current decommissioning strategy for the graphite-moderated reactors operated by Magnox Electric plc, Nuclear Electric Ltd and Scottish Nuclear Ltd is to delay dismantling and to initiate a monitored period of care and maintenance followed by a period of safe storage totaling up to 135 years. This philosophy has the considerable advantage of permitting the majority of radionuclides to decay, thereby minimising personnel dose during dismantling which itself will require far less complex remote-handling equipment. It also defers the disposal of the graphite and other components so that the provision of a deep land-based repository can be achieved. A comprehensive review of all relevant data on the chemical, physical and mechanical properties of the graphite and its potential reactions, including radioactivity transport, has been undertaken in order to demonstrate that there are no potential mechanisms which might lead to degradation of the core during the storage period. It is concluded that no significant experimental work is necessary to support the safe storage philosophy although, since the ingress of rainwater over long periods of time cannot be assumed incredible, a number of anomalies in chemical leaching rates may be worthy of re-examination. No other potential chemical reactions, such as the radiolytic formation of nitric acid leading to corrosion problems, are considered significant. (author)

  9. Beryllium and graphite performance in ITER during a disruption

    International Nuclear Information System (INIS)

    Hassanein, A.; Ehst, D.A.; Gahl, J.

    1993-09-01

    Plasma disruptions are considered one of the most limiting factors for successful operation of magnetic fusion reactors. During a disruption, a sharp, rapid release of energy strikes components such as the divertor or limiter plates. Severe surface erosion and melting of these components may then occur. The amount of material eroded from both ablation and melting is important to the reactor design and component lifetime. The anticipated performance of both beryllium and graphite as plasma-facing materials during such abnormal events is analyzed and compared. Recent experimental data obtained with both plasma guns and electron beams are carefully evaluated and compared to results of analytical modeling, including vapor shielding effect. Initial results from plasma gun experiments indicate that the Be erosion rate is about five times larger than that for a graphite material under the same disruption conditions. Key differences between simulation experiments and reactor disruption on the net erosion rate, and consequently on the lifetime of the divertor plate, are discussed in detail. The advantages and disadvantages of Be over graphite as a divertor plasma-facing material are discussed

  10. Beryllium and graphite performance in ITER during a disruption

    International Nuclear Information System (INIS)

    Hassanein, A.; Ehst, D.A.; Gahl, J.

    1994-01-01

    Plasma disruptions are considered one of the most limiting factors for successful operation of magnetic fusion reactors. During a disruption, a sharp, rapid release of energy strikes components such as the divertor or limiter plates. Severe surface erosion and melting of these components may then occur. The amount of material eroded from both ablation and melting is important to the reactor design and component lifetime. The anticipated performance of both beryllium and graphite as plasma-facing materials during such abnormal events is analyzed and compared. Recent experimental data obtained with both plasma guns and electron beams are carefully evaluated and compared to results of analytical modeling, including vapor shielding effect. Initial results from plasma gun experiments indicate that the Be erosion rate is about five times larger than that for a graphite material under the same disruption conditions. Key differences between simulation experiments and reactor disruption on the net erosion rate, and consequently on the lifetime of the divertor plate, are discussed in detail. The advantages and disadvantages of Be over graphite as a divertor plasma-facing material are discussed. ((orig.))

  11. Work function modifications of graphite surface via oxygen plasma treatment

    Science.gov (United States)

    Duch, J.; Kubisiak, P.; Adolfsson, K. H.; Hakkarainen, M.; Golda-Cepa, M.; Kotarba, A.

    2017-10-01

    The surface modification of graphite by oxygen plasma was investigated experimentally (X-ray diffraction, nanoparticle tracking analysis, laser desorption ionization mass spectrometry, thermogravimetry, water contact angle) and by molecular modelling (Density Functional Theory). Generation of surface functional groups (mainly sbnd OHsurf) leads to substantial changes in electrodonor properties and wettability gauged by work function and water contact angle, respectively. The invoked modifications were analyzed in terms of Helmholtz model taking into account the theoretically determined surface dipole moment of graphite-OHsurf system (μ = 2.71 D) and experimentally measured work function increase (from 0.75 to 1.02 eV) to determine the sbnd OH surface coverage (from 0.70 to 1.03 × 1014 groups cm-2). Since the plasma treatment was confined to the surface, the high thermal stability of the graphite material was preserved as revealed by the thermogravimetric analysis. The obtained results provide a suitable quantitative background for tuning the key operating parameters of carbon electrodes: electronic properties, interaction with water and thermal stability.

  12. 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.

  13. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 5: Graphite PIRTs

    International Nuclear Information System (INIS)

    Burchell, Timothy D.; Bratton, Rob; Marsden, Barry; Srinivasan, Makuteswara; Penfield, Scott; Mitchell, Mark; Windes, Will

    2008-01-01

    Here we report the outcome of the application of the Nuclear Regulatory Commission (NRC) Phenomena Identification and Ranking Table (PIRT) process to the issue of nuclear-grade graphite for the moderator and structural components of a next generation nuclear plant (NGNP), considering both routine (normal operation) and postulated accident conditions for the NGNP. The NGNP is assumed to be a modular high-temperature gas-cooled reactor (HTGR), either a gas-turbine modular helium reactor (GTMHR) version (a prismatic-core modular reactor (PMR)] or a pebble-bed modular reactor (PBMR) version (a pebble bed reactor (PBR)] design, with either a direct- or indirect-cycle gas turbine (Brayton cycle) system for electric power production, and an indirect-cycle component for hydrogen production. NGNP design options with a high-pressure steam generator (Rankine cycle) in the primary loop are not considered in this PIRT. This graphite PIRT was conducted in parallel with four other NRC PIRT activities, taking advantage of the relationships and overlaps in subject matter. The graphite PIRT panel identified numerous phenomena, five of which were ranked high importance-low knowledge. A further nine were ranked with high importance and medium knowledge rank. Two phenomena were ranked with medium importance and low knowledge, and a further 14 were ranked medium importance and medium knowledge rank. The last 12 phenomena were ranked with low importance and high knowledge rank (or similar combinations suggesting they have low priority). The ranking/scoring rationale for the reported graphite phenomena is discussed. Much has been learned about the behavior of graphite in reactor environments in the 60-plus years since the first graphite rectors went into service. The extensive list of references in the Bibliography is plainly testament to this fact. Our current knowledge base is well developed. Although data are lacking for the specific grades being considered for Generation IV (Gen IV

  14. A Survey On Mean Glandular Dose From Full-Field Digital Mammography Systems, Operate Using Mo/ Mo And W/Rh Target/ Filter Combinations

    International Nuclear Information System (INIS)

    Noriah Jamal; Siti Selina Abdul Hamid; Humairah Samad Cheung; Siti Kamariah Che Mohamed; Ellyda Muhammed Nordin; Radhiana Hassan; Rehir Dahalan

    2013-01-01

    We had conducted a survey on Mean Glandular Dose (MGD) from Full-Field Digital Mammography systems (FFDM) operate using Molybdenum/ Molybdenum (Mo/ Mo) and Tungsten/ Rhodium (W/ Rh) target/ filter combinations. A survey was carried out at two randomly selected mammography centres in Malaysia, namely National Cancer Society and International Islamic University of Malaysia. The first centre operates using a W/ Rh, while the second centre operates using an Mo/ Mo target/ filter combinations. On the basis of recorded information, data on mammographic views, MGD, age and Compressed Breast Thickness (CBT) were recorded for 100 patients, for each mammographic centre respectively. The MGD data were analyzed for variation with age group, with 5 years increment. The MGD data were also analyzed for variation with CBT, with 5 mm increment. We found that for both CC and MLO views, FFDM systems operated using Mo/ Mo and W/ Rh target/ filter combinations present the same trend on MGD. The average MGD decreases as age increases. While average MGD increases with the increasing of CBT. However, FFDM system operates using Mo/ Mo gives higher MGD as compared with FFDM system operates using W/ Rh. (author)

  15. 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.

  16. 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)

  17. NDA Position on the UK Management of Waste Graphite (December 2013)

    International Nuclear Information System (INIS)

    Norris, S.

    2016-01-01

    The purpose of this paper is to summarise a number of pieces of work that have been undertaken by the Nuclear Decommissioning Authority (NDA) to better understand the challenges of managing radioactive graphite wastes, these have led to an updated strategic position on graphite waste management. The updated strategic position takes into consideration Government’s response to Recommendation 8 from the Committee on Radioactive Waste Management’s (CoRWM), and provides the current NDA strategic position alongside circumstances where this should be reviewed. Two studies that provided input to this position are: 1. Operational Graphite Management Strategy: Credible and Preferred Options (Gate A & B); 2. The Long-term Management of Reactor Core Graphite Waste: Credible Options (Gate A). The paper highlights the key findings from the following work that has been undertaken to better inform this position: • A review by the NDA Radioactive Waste Management Directorate (RWMD)1 of the current baseline for managing radioactive graphite in England and Wales of geological disposal. The review identified some areas for optimisation and provided clarification on some aspects of the baseline e.g. the assumed ‘footprint’ of graphite wastes for a future Geological Disposal Facility (GDF). • Investigations into suitability of near-surface disposal options for graphite wastes. This included a review of the Low Level Waste Repository (LLWR) Ltd's new Environmental Safety Case (ESC) to assess the potential for graphite disposal and a feasibility study into a near-surface disposal facility for Higher Activity Waste (HAW) graphite at the Hunterston A site. • Continued monitoring of potential future treatment options. • Detailed characterisation work under the NDA’s Direct Research Portfolio using computer modelling and sample analysis to better understand any limitations of the current inventory data for graphite wastes. • Graphite behaviour work under the NDA

  18. Graphite epoxy composite degradation by space radiation

    International Nuclear Information System (INIS)

    Taheri, M.; Sandquist, G.M.; Slaughter, D.M.; Bennion, J.

    1991-01-01

    The radiation environment in space is a critical consideration for successful operation in space. All manned space missions with a duration of more than a few days are subjected to elevated ionizing radiation exposures, which are a threat to both personnel and structures in space. The increasing demands for high-performance materials as structural components in the aerospace, aircraft, and defense industries have led to the development of materials such as graphite fiber-reinforced, epoxy resin matrix composites (Gr/Ep). These materials provide important advantages over conventional structural materials, such as ultrahigh specific strength, enhanced specific moduli, and better fatigue resistance. The fact that most advanced composite materials under cyclic fatigue loading evidence little or no observable crack growth prior to rapid fracture suggests that for fail-safe considerations of parts subject to catastrophic failure, a detailed evaluation of radiation damage from very energetic particle is crucial. The Gr/Ep components are believed to suffer severe degradation in space due to highly penetrating secondary radiation, mainly from neutrons and protons. Investigation into the performance and stability of Gr/Ep materials are planned

  19. Pyrolytic Graphite as a Selective Neutron Filter

    International Nuclear Information System (INIS)

    Adib, M.; Habib, N.; Fathalla, M.

    2006-01-01

    The transmission of neutrons through pyrolytic graphite (PG) crystals, set at different angles with respect to incident beam, were calculated using an additive formula. A computer program HOPG was developed to provide the required calculation. An overall agreement between the calculated neutron transmissions through a slab of 1,85 mm thick PG crystal with an angular spread of c-axes of 0,4 degree, set at different angles to the incident beam, and the available experimental ones in the wavelength range from (0,02 to 1,4) nm were obtained. A feasibility study for use of PG crystal as an efficient second-order neutron filter is detailed in terms of crystal thickness, angular spread of c-axes and its operation with respect to the neutron beam. It was shown that a PG crystal with an angular spread of c-axes and its orientation with respect to the neutron beam. It was shown that a PG crystal with an angular spread of 0,8 degree is sufficient for optimum scattering of second-order neutrons in the wavelength band (0,384-0,183) nm, by adjusting the filter in an appropriate orientation

  20. 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

  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. 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...

  3. 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

  4. 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.)

  5. 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

  6. 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.

  7. 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

  8. 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.

  9. 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...

  10. 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

  11. INTERACTION OF LASER RADIATION WITH MATTER: Influence of a target on operation of a pulsed CO2 laser emitting microsecond pulses

    Science.gov (United States)

    Baranov, V. Yu; Dolgov, V. A.; Malyuta, D. D.; Mezhevov, V. S.; Semak, V. V.

    1987-12-01

    The profile of pulses emitted by a TEA CO2 laser with an unstable resonator changed as a result of interaction of laser radiation with the surface of a metal in the presence of a breakdown plasma. This influence of a target on laser operation and its possible applications in laser processing of materials are analyzed.

  12. Investigating the effects of stress on the pore structures of nuclear grade graphites

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Joshua E.L., E-mail: joshua.taylor@postgrad.manchester.ac.uk; Hall, Graham N., E-mail: graham.n.hall@manchester.ac.uk; Mummery, Paul M., E-mail: paul.m.mummery@manchester.ac.uk

    2016-03-15

    Graphite is used as a moderating material and as a structural component in a number of current generation nuclear reactors. During reactor operation stresses develop in the graphite components, causing them to deform. It is important to understand how the microstructure of graphite affects the material's response to these stresses. A series of experiments were performed to investigate how the pore structures of Pile Grade A and Gilsocarbon graphites respond to loading stresses. A compression rig was used to simulate the build-up of operational stresses in graphite components, and a confocal laser microscope was used to study variation of a number of important pore properties. Values of elastic modulus and Poisson's ratio were calculated and compared to existing literature to confirm the validity of the experimental techniques. Mean pore areas were observed to decrease linearly with increasing applied load, mean pore eccentricity increased linearly, and a small amount of clockwise pore rotation was observed. The response to build-up of stresses was dependent on the orientation of the pores and basal planes and the shapes of the pores with respect to the loading axis. It was proposed that pore closure and pore reorientation were competing processes. Pore separation was quantified using ‘nearest neighbour’ and Voronoi techniques, and non-pore regions were found to shrink linearly with increasing applied load. - Highlights: • Effects of stress on pore structures of Gilsocarbon and PGA graphites were studied. • Application of a compressive load was used to generate stresses in graphite. • Inverse linear relationship between stress and pore area was observed. • Mean pore eccentricity increased, clockwise pore rotation observed. • Separation of pores quantified using Voronoi and ‘nearest-neighbour’ methods.

  13. 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)

  14. 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

  15. 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)

  16. Structure and functionality of bromine doped graphite.

    Science.gov (United States)

    Hamdan, Rashid; Kemper, A F; Cao, Chao; Cheng, H P

    2013-04-28

    First-principles calculations are used to study the enhanced in-plane conductivity observed experimentally in Br-doped graphite, and to study the effect of external stress on the structure and functionality of such systems. The model used in the numerical calculations is that of stage two doped graphite. The band structure near the Fermi surface of the doped systems with different bromine concentrations is compared to that of pure graphite, and the charge transfer between carbon and bromine atoms is analyzed to understand the conductivity change along different high symmetry directions. Our calculations show that, for large interlayer separation between doped graphite layers, bromine is stable in the molecular form (Br2). However, with increased compression (decreased layer-layer separation) Br2 molecules tend to dissociate. While in both forms, bromine is an electron acceptor. The charge exchange between the graphite layers and Br atoms is higher than that with Br2 molecules. Electron transfer to the Br atoms increases the number of hole carriers in the graphite sheets, resulting in an increase of conductivity.

  17. Characterization of radiation damage induced by swift heavy ions in graphite

    Energy Technology Data Exchange (ETDEWEB)

    Hubert, Christian

    2016-05-15

    irradiation temperatures above 200 C damage formation is mitigated due to defect annealing. Thus a controlled temperature of accelerator components is desirable in order to increase the lifetime. This thesis contributes to a better understanding of radiation damage in swift heavy ion-exposed graphite with the aim to optimize the design of beam catchers and production targets for secondary ion beams for the Super Fragment Separator (Super-FRS) at FAIR. Moreover, the results of this work provide important input data for simulations to describe the beam response and lifetime of high-dose exposed critical accelerator components.

  18. Energy-momentum density of graphite by electron-momentum spectroscopy

    International Nuclear Information System (INIS)

    Vos, M.; Fang, Z.; Canney, S.; Kheifets, A.; McCarthy, I.E.; Weigold, E.

    1996-11-01

    The energy-resolved electron momentum density of graphite has been measured along a series of well-defined directions using electron momentum spectroscopy (EMS). This is the first measurement of this kind performed on a single-crystal target with a thoroughly controlled orientation which clearly demonstrates the different nature of the σ and π bands in graphite. Good agreement between the calculated density and the measured one is found, further establishing that fact that EMS yields more direct and complete information on the valence electronic structure that any other method. 12 refs., 2 figs

  19. Neutron irradiation damage of nuclear graphite studied by high-resolution transmission electron microscopy and Raman spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Krishna, R. [Dalton Cumbrian Facility, Dalton Nuclear Institute, The University of Manchester, Westlakes Science & Technology Park, Moor Row, Whitehaven, Cumbria, CA24 3HA (United Kingdom); Jones, A.N., E-mail: Abbie.Jones@manchester.ac.uk [Nuclear Graphite Research Group, School of MACE, The University of Manchester, Manchester, M13 9PL (United Kingdom); McDermott, L.; Marsden, B.J. [Nuclear Graphite Research Group, School of MACE, The University of Manchester, Manchester, M13 9PL (United Kingdom)

    2015-12-15

    Nuclear graphite components are produced from polycrystalline artificial graphite manufacture from a binder and filler coke with approximately 20% porosity. During the operational lifetime, nuclear graphite moderator components are subjected to fast neutron irradiation which contributes to the change of material and physical properties such as thermal expansion co-efficient, young's modulus and dimensional change. These changes are directly driven by irradiation-induced changes to the crystal structure as reflected through the bulk microstructure. It is therefore of critical importance that these irradiation changes and there implication on component property changes are fully understood. This work examines a range of irradiated graphite samples removed from the British Experimental Pile Zero (BEPO) reactor; a low temperature, low fluence, air-cooled Materials Test Reactor which operated in the UK. Raman spectroscopy and high-resolution transmission electron microscopy (HRTEM) have been employed to characterise the effect of increased irradiation fluence on graphite microstructure and understand low temperature irradiation damage processes. HRTEM confirms the structural damage of the crystal lattice caused by irradiation attributed to a high number of defects generation with the accumulation of dislocation interactions at nano-scale range. Irradiation-induced crystal defects, lattice parameters and crystallite size compared to virgin nuclear graphite are characterised using selected area diffraction (SAD) patterns in TEM and Raman Spectroscopy. The consolidated ‘D’peak in the Raman spectra confirms the formation of in-plane point defects and reflected as disordered regions in the lattice. The reduced intensity and broadened peaks of ‘G’ and ‘D’ in the Raman and HRTEM results confirm the appearance of turbulence and disordering of the basal planes whilst maintaining their coherent layered graphite structure. - Highlights: • Irradiated graphite

  20. Neutron irradiation damage of nuclear graphite studied by high-resolution transmission electron microscopy and Raman spectroscopy

    International Nuclear Information System (INIS)

    Krishna, R.; Jones, A.N.; McDermott, L.; Marsden, B.J.

    2015-01-01

    Nuclear graphite components are produced from polycrystalline artificial graphite manufacture from a binder and filler coke with approximately 20% porosity. During the operational lifetime, nuclear graphite moderator components are subjected to fast neutron irradiation which contributes to the change of material and physical properties such as thermal expansion co-efficient, young's modulus and dimensional change. These changes are directly driven by irradiation-induced changes to the crystal structure as reflected through the bulk microstructure. It is therefore of critical importance that these irradiation changes and there implication on component property changes are fully understood. This work examines a range of irradiated graphite samples removed from the British Experimental Pile Zero (BEPO) reactor; a low temperature, low fluence, air-cooled Materials Test Reactor which operated in the UK. Raman spectroscopy and high-resolution transmission electron microscopy (HRTEM) have been employed to characterise the effect of increased irradiation fluence on graphite microstructure and understand low temperature irradiation damage processes. HRTEM confirms the structural damage of the crystal lattice caused by irradiation attributed to a high number of defects generation with the accumulation of dislocation interactions at nano-scale range. Irradiation-induced crystal defects, lattice parameters and crystallite size compared to virgin nuclear graphite are characterised using selected area diffraction (SAD) patterns in TEM and Raman Spectroscopy. The consolidated ‘D’peak in the Raman spectra confirms the formation of in-plane point defects and reflected as disordered regions in the lattice. The reduced intensity and broadened peaks of ‘G’ and ‘D’ in the Raman and HRTEM results confirm the appearance of turbulence and disordering of the basal planes whilst maintaining their coherent layered graphite structure. - Highlights: • Irradiated graphite exhibits

  1. Graphite-modified cooling ceiling. A comparative evaluation of the performance; Graphitmodifizierte Kuehldecken. Eine vergleichende Bewertung zur Leistungsfaehigkeit

    Energy Technology Data Exchange (ETDEWEB)

    Hauser, Gerd [Technische Univ. Muenchen (Germany); Fraunhofer-Institut fuer Bauphysik, Stuttgart (Germany); Hausladen, Gerhard [Technische Univ. Muenchen (Germany); Ingenieurbuero Hausladen GmbH, Muenchen (Germany); Jacobsen, Cornelia; Hutter, Christina; Hanusch, Christoph [Ingenieurbuero Hausladen GmbH, Muenchen (Germany); Schlitzberger, Stephan [Ingenieurbuero fuer Bauphysik Prof. Dr. Hauser GmbH, Kassel (Germany)

    2013-03-01

    In most cases, the utilization of graphite-modified building materials enhances their thermal conductivity. Thus, chilled ceiling systems achieve an enhanced thermal absorption in comparison to commercial chilled ceiling systems. For this, a simulation for the estimation of the thermal comfort as well as for the quantification of the impacts on the demand of electricity for the refrigeration is performed. The enhanced cooling capacity of the graphite-modified cooling ceiling 'Ecophit' affects the operation positively.

  2. AA antiproton production target

    CERN Multimedia

    CERN PhotoLab

    1979-01-01

    The first version of the antiproton production target was a tungsten rod, 11 cm long and 3 mm in diameter. The rod was embedded in graphite, pressure-seated into an outer casing of stainless steel. At the entrance to the target assembly was a scintillator screen, imprinted with circles every 5 mm in radius, which allowed to precisely aim the 26 GeV high-intensity proton beam from the PS onto the centre of the target rod. The scintillator screen was a 1 mm thick plate of Cr-doped alumina. See also 7903034 and 7905091.

  3. Production of nuclear graphite in France; Production de graphite nucleaire en France

    Energy Technology Data Exchange (ETDEWEB)

    Legendre, P; Mondet, L [Societe Pechiney, 74 - Chedde (France); Arragon, Ph; Cornuault, P; Gueron, J; Hering, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    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.) [French] Le graphite destine a la construction des reacteurs est obtenu par le procede usuel: confection d'une pate a partir de coke de petrole et de brai, cuisson de cette pate (au four electrique) puis graphitation du produit cuit, egalement par chauffage electrique. L'usage du transport pneumatique et le controle des conditions cuisson et de graphitation ont permit d'augmenter la production de graphite nucleaire ainsi que de mieux controler ses proprietes physiques et mecaniques et de reduire au minimum les souillures accidentelles. (M.B.)

  4. Graphite matrix materials for nuclear waste isolation

    International Nuclear Information System (INIS)

    Morgan, W.C.

    1981-06-01

    At low temperatures, graphites are chemically inert to all but the strongest oxidizing agents. The raw materials from which artificial graphites are produced are plentiful and inexpensive. Morover, the physical properties of artificial graphites can be varied over a very wide range by the choice of raw materials and manufacturing processes. Manufacturing processes are reviewed herein, with primary emphasis on those processes which might be used to produce a graphite matrix for the waste forms. The approach, recommended herein, involves the low-temperature compaction of a finely ground powder produced from graphitized petroleum coke. The resultant compacts should have fairly good strength, low permeability to both liquids and gases, and anisotropic physical properties. In particular, the anisotropy of the thermal expansion coefficients and the thermal conductivity should be advantageous for this application. With two possible exceptions, the graphite matrix appears to be superior to the metal alloy matrices which have been recommended in prior studies. The two possible exceptions are the requirements on strength and permeability; both requirements will be strongly influenced by the containment design, including the choice of materials and the waste form, of the multibarrier package. Various methods for increasing the strength, and for decreasing the permeability of the matrix, are reviewed and discussed in the sections in Incorporation of Other Materials and Elimination of Porosity. However, it would be premature to recommend a particular process until the overall multi-barrier design is better defined. It is recommended that increased emphasis be placed on further development of the low-temperature compacted graphite matrix concept

  5. Method of producing exfoliated graphite, flexible graphite, and nano-scaled graphene platelets

    Science.gov (United States)

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

    2010-11-02

    The present invention provides a method of exfoliating a layered material (e.g., graphite and graphite oxide) to produce nano-scaled platelets having a thickness smaller than 100 nm, typically smaller than 10 nm. The method comprises (a) dispersing particles of graphite, graphite oxide, or a non-graphite laminar compound in a liquid medium containing therein a surfactant or dispersing agent to obtain a stable suspension or slurry; and (b) exposing the suspension or slurry to ultrasonic waves at an energy level for a sufficient length of time to produce separated nano-scaled platelets. The nano-scaled platelets are candidate reinforcement fillers for polymer nanocomposites. Nano-scaled graphene platelets are much lower-cost alternatives to carbon nano-tubes or carbon nano-fibers.

  6. EEL Calculations and Measurements of Graphite and Graphitic-CNx Core-Losses

    International Nuclear Information System (INIS)

    Seepujak, A; Bangert, U; Harvey, A J; Blank, V D; Kulnitskiy, B A; Batov, D V

    2006-01-01

    Core EEL spectra of MWCNTs (multi-wall carbon nanotubes) grown in a nitrogen atmosphere were acquired utilising a dedicated STEM equipped with a Gatan Enfina system. Splitting of the carbon K-edge π* resonance into two peaks provided evidence of two nondegenerate carbon bonding states. In order to confirm the presence of a CN x bonding state, a full-potential linearised augmented plane-wave method was utilised to simulate core EEL spectra of graphite and graphitic-CN x compounds. The simulations confirmed splitting of the carbon K-edge π* resonance in graphitic-CN x materials, with the pristine graphite π* resonance remaining unsplit. The simulations also confirmed the increasing degree of amorphicity with higher concentrations (25%) of substitutional nitrogen in graphite

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  8. Surface impurity removal from DIII-D graphite tiles by boron carbide grit blasting

    International Nuclear Information System (INIS)

    Lee, R.L.; Hollerbach, M.A.; Holtrop, K.L.; Kellman, A.G.; Taylor, P.L.; West, W.P.

    1993-11-01

    During the latter half of 1992, the DIII-D tokamak at General Atomics (GA) underwent several modifications of its interior. One of the major tasks involved the removal of accumulated metallic impurities from the surface of the graphite tiles used to line the plasma facing surfaces inside of the tokamak. Approximately 1500 graphite tiles and 100 boron nitride tiles from the tokamak were cleaned to remove the metallic impurities. The cleaning process consisted of several steps: the removed graphite tiles were permanently marked, surface blasted using boron carbide (B 4 C) grit media (approximately 37 μm. diam.), ultrasonically cleaned in ethanol to remove loose dust, and outgassed at 1000 degrees C. Tests were done using, graphite samples and different grit blaster settings to determine the optimum propellant and abrasive media pressures to remove a graphite layer approximately 40-50 μm deep and yet produce a reasonably smooth finish. EDX measurements revealed that the blasting technique reduced the surface Ni, Cr, and Fe impurity levels to those of virgin graphite. In addition to the surface impurity removal, tritium monitoring was performed throughout the cleaning process. A bubbler system was set up to monitor the tritium level in the exhaust gas from the grit blaster unit. Surface wipes were also performed on over 10% of the tiles. Typical surface tritium concentrations of the tiles were reduced from about 500 dpm/100 cm 2 to less than 80 dpm/100 cm 2 following the cleaning. This tile conditioning, and the installation of additional graphite tiles to cover a high fraction of the metallic plasma facing surfaces, has substantially reduced metallic impurities in the plasma discharges which has allowed rapid recovery from a seven-month machine opening and regimes of enhanced plasma energy confinement to be more readily obtained. Safety issues concerning blaster operator exposure to carcinogenic metals and radioactive tritium will also be addressed

  9. Gas transport in graphitic materials

    International Nuclear Information System (INIS)

    Hoinkis, E.

    1995-02-01

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

  10. A free-jet Hg target operating in a high magnetic field intersecting a high-power proton beam

    Science.gov (United States)

    Van Graves; Spampinato, Philip; Gabriel, Tony; Kirk, Harold; Simos, Nicholas; Tsang, Thomas; McDonald, Kirk; Peter Titus; Fabich, Adrian; Haseroth, Helmut; Lettry, Jacques

    2006-06-01

    A proof-of-principal experiment to investigate the interaction of a proton beam, high magnetic field, and high- Z target is planned to take place at CERN in early 2007. This experiment is part of the Muon Collider Collaboration, with participants from Brookhaven National Laboratory, Princeton University, Massachusetts Institute Of Technology, European Organization for Nuclear Research-CERN, Rutherford Appleton Laboratory, and Oak Ridge National Laboratory. An unconstrained mercury jet target system that interacts with a high power (1 MW) proton beam in a high magnetic field (15 T) is being designed. The Hg jet diameter is 1-cm with a velocity up to 20 m/s. A laser optical diagnostic system will be incorporated into the target design to permit observation of the dispersal of the jet resulting from interaction with a 24 GeV proton beam with up to 20×10 12 ppp. The target system includes instruments for sensing mercury vapor, temperature, flow rate, and sump tank level, and the means to position the jet relative to the magnetic axis of a solenoid and the proton beam. The design considerations for the system include all issues dealing with safely handling approximately 23 l of Hg, transporting the target system and the mercury to CERN, decommissioning the experiment, and returning the mildly activated equipment and Hg to the US.

  11. Influence of Metal-Coated Graphite Powders on Microstructure and Properties of the Bronze-Matrix/Graphite Composites

    Science.gov (United States)

    Zhao, Jian-hua; Li, Pu; Tang, Qi; Zhang, Yan-qing; He, Jian-sheng; He, Ke

    2017-02-01

    In this study, the bronze-matrix/x-graphite (x = 0, 1, 3 and 5%) composites were fabricated by powder metallurgy route by using Cu-coated graphite, Ni-coated graphite and pure graphite, respectively. The microstructure, mechanical properties and corrosive behaviors of bronze/Cu-coated-graphite (BCG), bronze/Ni-coated-graphite (BNG) and bronze/pure-graphite (BPG) were characterized and investigated. Results show that the Cu-coated and Ni-coated graphite could definitely increase the bonding quality between the bronze matrix and graphite. In general, with the increase in graphite content in bronze-matrix/graphite composites, the friction coefficients, ultimate density and wear rates of BPG, BCG and BNG composites all went down. However, the Vickers microhardness of the BNG composite would increase as the graphite content increased, which was contrary to the BPG and BCG composites. When the graphite content was 3%, the friction coefficient of BNG composite was more stable than that of BCG and BPG composites, indicating that BNG composite had a better tribological performance than the others. Under all the values of applied loads (10, 20, 40 and 60N), the BCG and BNG composites exhibited a lower wear rate than BPG composite. What is more, the existence of nickel in graphite powders could effectively improve the corrosion resistance of the BNG composite.

  12. Computational prediction of dust production in graphite moderated pebble bed reactors

    Science.gov (United States)

    Rostamian, Maziar

    The scope of the work reported here, which is the computational study of graphite wear behavior, supports the Nuclear Engineering University Programs project "Experimental Study and Computational Simulations of Key Pebble Bed Thermomechanics Issues for Design and Safety" funded by the US Department of Energy. In this work, modeling and simulating the contact mechanics, as anticipated in a PBR configuration, is carried out for the purpose of assessing the amount of dust generated during a full power operation year of a PBR. A methodology that encompasses finite element analysis (FEA) and micromechanics of wear is developed to address the issue of dust production and its quantification. Particularly, the phenomenon of wear and change of its rate with sliding length is the main focus of this dissertation. This work studies the wear properties of graphite by simulating pebble motion and interactions of a specific type of nuclear grade graphite, IG-11. This study consists of two perspectives: macroscale stress analysis and microscale analysis of wear mechanisms. The first is a set of FEA simulations considering pebble-pebble frictional contact. In these simulations, the mass of generated graphite particulates due to frictional contact is calculated by incorporating FEA results into Archard's equation, which is a linear correlation between wear mass and wear length. However, the experimental data by Johnson, University of Idaho, revealed that the wear rate of graphite decreases with sliding length. This is because the surfaces of the graphite pebbles become smoother over time, which results in a gradual decrease in wear rate. In order to address the change in wear rate, a more detailed analysis of wear mechanisms at room temperature is presented. In this microscale study, the wear behavior of graphite at the asperity level is studied by simulating the contact between asperities of facing surfaces. By introducing the effect of asperity removal on wear rate, a nonlinear

  13. Interim results: fines recycle testing using the 4-inch diameter primary graphite burner

    International Nuclear Information System (INIS)

    Palmer, W.B.

    1975-05-01

    The results of twenty-two HTGR primary burner runs in which graphite fines were recycled pneumatically to the 4-inch diameter pilot-plant primary fluidized-bed burner are described. The result of the tests showed that zero fines accumulation can easily be achieved while operating at plant equivalent burn rates. (U.S.)

  14. 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

  15. 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

  16. The mechanical behavior and reliability prediction of the HTR graphite component at various temperature and neutron dose ranges

    International Nuclear Information System (INIS)

    Fang, Xiang; Yu, Suyuan; Wang, Haitao; Li, Chenfeng

    2014-01-01

    Highlights: • The mechanical behavior of graphite component in HTRs under high temperature and neutron irradiation conditions is simulated. • The computational process of mechanical analysis is introduced. • Deformation, stresses and failure probability of the graphite component are obtained and discussed. • Various temperature and neutron dose ranges are selected in order to investigate the effect of in-core conditions on the results. - Abstract: In a pebble-bed high temperature gas-cooled reactor (HTR), nuclear graphite serves as the main structural material of the side reflectors. The reactor core is made up of a large number of graphite bricks. In the normal operation case of the reactor, the maximum temperature of the helium coolant commonly reaches about 750 °C. After around 30 years’ full power operation, the peak value of in-core fast neutron cumulative dose reaches to 1 × 10 22 n cm −2 (EDN). Such high temperature and neutron irradiation strongly impact the behavior of graphite component, causing obvious deformation. The temperature and neutron dose are unevenly distributed inside a graphite brick, resulting in stress concentrations. The deformation and stress concentration can both greatly affect safety and reliability of the graphite component. In addition, most of the graphite properties (such as Young's modulus and coefficient of thermal expansion) change remarkably under high temperature and neutron irradiations. The irradiation-induced creep also plays a very important role during the whole process, and provides a significant impact on the stress accumulation. In order to simulate the behavior of graphite component under various in-core conditions, all of the above factors must be considered carefully. In this paper, the deformation, stress distribution and failure probability of a side graphite component are studied at various temperature points and neutron dose levels. 400 °C, 500 °C, 600 °C and 750 °C are selected as the

  17. Hydrogen adsorption on and solubility in graphites

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  18. Irradiation-induced amorphization process in graphite

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Hiroaki [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment

    1996-04-01

    Effects of the element process of irradiation damage on irradiation-induced amorphization processes of graphite was studied. High orientation thermal decomposed graphite was cut about 100 nm width and used as samples. The irradiation experiments are carried out under the conditions of electronic energy of 100-400 KeV, ion energy of 200-600 KeV, ionic species Xe, Ar, Ne, C and He and the irradiation temperature at from room temperature to 900 K. The critical dose ({phi}a) increases exponentially with increasing irradiation temperature. The displacement threshold energy of graphite on c-axis direction was 27 eV and {phi}a{sup e} = 0.5 dpa. dpa is the average number of displacement to atom. The critical dose of ion irradiation ({phi}a{sup i}) was 0.2 dpa at room temperature, and amorphous graphite was produced by less than half of dose of electronic irradiation. Amorphization of graphite depending upon temperature is discussed. (S.Y.)

  19. Interactions of hydrogen with graphite at low pressure and elevated temperature

    International Nuclear Information System (INIS)

    Hoinkis, E.

    1991-03-01

    The plasma facing components of the vacuum chamber for thermonuclear fusion experiments are clad with graphite. Recycling of gases affects the plasma properties, and the tritium quantity accumulated in the graphite during the operation of Tokamaks with DT must be known. An adsorption isotherm for deuterium on the nuclear grade graphitic Matrix A3-3 was measured by using a volumetric method at 1173 K at pressures c = 2.5 eV/D 2 using Fowler's equation and isotherms were calculated for this E c value. These isotherms predict saturation of the adsorption sites in graphite at T D2 > 0.1 Pa. At T > 1173 K and P D2 -2 Pa the adsorbed quantity is less than 1% of the saturation level. The release kinetics of deuterium was measured at temperatures uo to 2000 K. D 2 desorption commenced at 1170 K. The maximum of the release rate is observed at T p = 1770 K. A Lennard-Jones potential energy diagram was calculated, which suggests a C-H bond energy E b ≅ 3.4 eV/D and an activation energy of desorption E d ≅ 4 eV/D 2 . The partial pressures of hydrocarbons C n ≤3 H m in equilibrium with graphite were calculated. At total pressures -2 Pa the partial pressures of these hydrocarbons are less than 10 -7 Pa in the temperature range 600-1500 K. (orig./MM)

  20. Formation mechanism of the graphite-rich protective layer in blast furnace hearths

    Science.gov (United States)

    Jiao, Ke-xin; Zhang, Jian-liang; Liu, Zheng-jian; Liu, Feng; Liang, Li-sheng

    2016-01-01

    A long campaign life of blast furnaces is heavily linked to the existence of a protective layer in their hearths. In this work, we conducted dissection studies and investigated damage in blast furnace hearths to estimate the formation mechanism of the protective layer. The results illustrate that a significant amount of graphite phase was trapped within the hearth protective layer. Furthermore, on the basis of the thermodynamic and kinetic calculations of the graphite precipitation process, a precipitation potential index related to the formation of the graphite-rich protective layer was proposed to characterize the formation ability of this layer. We determined that, under normal operating conditions, the precipitation of graphite phase from hot metal was thermodynamically possible. Among elements that exist in hot metal, C, Si, and P favor graphite precipitation, whereas Mn and Cr inhibit this process. Moreover, at the same hot-face temperature, an increase of carbon concentration in hot metal can shorten the precipitation time. Finally, the results suggest that measures such as reducing the hot-face temperature and increasing the degree of carbon saturation in hot metal are critically important to improve the precipitation potential index.

  1. Top-Emission Organic Light Emitting Diode Fabrication Using High Dissipation Graphite Substrate

    Directory of Open Access Journals (Sweden)

    Yu-Sheng Tsai

    2014-01-01

    Full Text Available This study uses a synthetic graphite fiber as the heat dissipation substrate for top-emission organic light emitting diode (TEOLED to reduce the impact from joule heat. UV glue (YCD91 was spin coated onto the substrate as the insulation layer. The TEOLED structure is (glass; copper; graphite substrate/YCD91 glue/Al/Au/EHI608/TAPC/Alq3/LiF/Al/Ag. The proposed graphite fiber substrate presents better luminous performance compared with glass and copper substrate devices with luminance of 3055 cd/m2 and current efficiency of 6.11 cd/A at 50 mA/cm2. When lighting period of different substrates TEOLED, the substrate case back temperature was observed using different lighting periods. A glass substrate element operating from 5 to 25 seconds at 3000 cd/m2 luminance produced a temperature rate of 1.207°C/sec. Under 4000 cd/m2 luminance the copper and graphite substrate temperature rates were 0.125°C/sec and 0.088°C/sec. Graphite component lifetime was determined to be 1.875 times higher than the glass components and 1.125 times higher than that of copper.

  2. Guidelines for target volume definition in post-operative radiotherapy for prostate cancer, on behalf of the EORTC Radiation Oncology Group

    International Nuclear Information System (INIS)

    Poortmans, Philip; Bossi, Alberto; Vandeputte, Katia; Bosset, Mathieu; Miralbell, Raymond; Maingon, Philippe; Boehmer, Dirk; Budiharto, Tom; Symon, Zvi; Bergh, Alfons C.M. van den; Scrase, Christopher; Poppel, Hendrik van; Bolla, Michel

    2007-01-01

    The appropriate application of 3-D conformal radiotherapy, intensity modulated radiotherapy or image guided radiotherapy for patients undergoing post-operative radiotherapy for prostate cancer requires a standardisation of the target volume definition and delineation as well as standardisation of the clinical quality assurance procedures. Recommendations for this are presented on behalf of the European Organisation for Research and Treatment of Cancer (EORTC) Radiation Oncology Group and in addition to the already published guidelines for radiotherapy as the primary treatment

  3. Thermal fatigue of refractory metal/graphite composites for fusion applications

    International Nuclear Information System (INIS)

    Smid, I; Nickel, H.; Kny, E.; Reheis, N.

    1995-01-01

    Reactor grade graphite and molybdenum (TZM) were brazed with different high temperature brazes. The resulting composite tiles had a size of 50 mm x 50 mm with a graphite thickness of 10 mm and a TZM thickness of 5 mm. The brazed composites have been tested in electron beam simulation for their thermal fatigue properties. The parameters of these tests were chosen to match NET design specifications for normal operation and 'slow' peak energy deposition. The resulting damages and microstructural changes on the graphites and the brazes are discussed. Additional information is supplied on X-ray diffraction data proving the presence of different phases in the brazes. Finally the influence of a hydrogen plasma on the adaptability of the investigated brazes in fusion devices is discussed. (author)

  4. Quality assurance for the IAEA International Database on Irradiated Nuclear Graphite Properties

    International Nuclear Information System (INIS)

    Wickham, A.J.; Humbert, D.

    2006-06-01

    Consideration has been given to the process of Quality Assurance applied to data entered into current versions of the IAEA International Database on Irradiated Nuclear Graphite Properties. Originally conceived simply as a means of collecting and preserving data on irradiation experiments and reactor operation, the data are increasingly being utilised for the preparation of safety arguments and in the design of new graphites for forthcoming generations of graphite-moderated plant. Under these circumstances, regulatory agencies require assurances that the data are of appropriate accuracy and correctly transcribed, that obvious errors in the original documentation are either highlighted or corrected, etc., before they are prepared to accept analyses built upon these data. The processes employed in the data transcription are described in this document, and proposals are made for the categorisation of data and for error reporting by Database users. (author)

  5. Dislocation density and graphitization of diamond crystals

    International Nuclear Information System (INIS)

    Pantea, C.; Voronin, G.A.; Zerda, T.W.; Gubicza, J.; Ungar, T.

    2002-01-01

    Two sets of diamond specimens compressed at 2 GPa at temperatures varying between 1060 K and 1760 K were prepared; one in which graphitization was promoted by the presence of water and another in which graphitization of diamond was practically absent. X-ray diffraction peak profiles of both sets were analyzed for the microstructure by using the modified Williamson-Hall method and by fitting the Fourier coefficients of the measured profiles by theoretical functions for crystallite size and lattice strain. The procedures determined mean size and size distribution of crystallites as well as the density and the character of the dislocations. The same experimental conditions resulted in different microstructures for the two sets of samples. They were explained in terms of hydrostatic conditions present in the graphitized samples

  6. Capacitive behavior of highly-oxidized graphite

    Science.gov (United States)

    Ciszewski, Mateusz; Mianowski, Andrzej

    2014-09-01

    Capacitive behavior of a highly-oxidized graphite is presented in this paper. The graphite oxide was synthesized using an oxidizing mixture of potassium chlorate and concentrated fuming nitric acid. As-oxidized graphite was quantitatively and qualitatively analyzed with respect to the oxygen content and the species of oxygen-containing groups. Electrochemical measurements were performed in a two-electrode symmetric cell using KOH electrolyte. It was shown that prolonged oxidation causes an increase in the oxygen content while the interlayer distance remains constant. Specific capacitance increased with oxygen content in the electrode as a result of pseudo-capacitive effects, from 0.47 to 0.54 F/g for a scan rate of 20 mV/s and 0.67 to 1.15 F/g for a scan rate of 5 mV/s. Better cyclability was observed for the electrode with a higher oxygen amount.

  7. Reactivity of lithium exposed graphite surface

    International Nuclear Information System (INIS)

    Harilal, S.S.; Allain, J.P.; Hassanein, A.; Hendricks, M.R.; Nieto-Perez, M.

    2009-01-01

    Lithium as a plasma-facing component has many attractive features in fusion devices. We investigated chemical properties of the lithiated graphite surfaces during deposition using X-ray photoelectron spectroscopy and low-energy ion scattering spectroscopy. In this study we try to address some of the known issues during lithium deposition, viz., the chemical state of lithium on graphite substrate, oxide layer formation mechanisms, Li passivation effects over time, and chemical change during exposure of the sample to ambient air. X-ray photoelectron studies indicate changes in the chemical composition with various thickness of lithium on graphite during deposition. An oxide layer formation is noticed during lithium deposition even though all the experiments were performed in ultrahigh vacuum. The metal oxide is immediately transformed into carbonate when the deposited sample is exposed to air.

  8. Reduced graphite oxide in supercapacitor electrodes.

    Science.gov (United States)

    Lobato, Belén; Vretenár, Viliam; Kotrusz, Peter; Hulman, Martin; Centeno, Teresa A

    2015-05-15

    The current energy needs have put the focus on highly efficient energy storage systems such as supercapacitors. At present, much attention focuses on graphene-like materials as promising supercapacitor electrodes. Here we show that reduced graphite oxide offers a very interesting potential. Materials obtained by oxidation of natural graphite and subsequent sonication and reduction by hydrazine achieve specific capacitances as high as 170 F/g in H2SO4 and 84F/g in (C2H5)4NBF4/acetonitrile. Although the particle size of the raw graphite has no significant effect on the physico-chemical characteristics of the reduced materials, that exfoliated from smaller particles (materials may suffer from a drop in their specific surface area upon fabrication of electrodes with features of the existing commercial devices. This should be taken into account for a reliable interpretation of their performance in supercapacitors. Copyright © 2015 Elsevier Inc. All rights reserved.

  9. Graphite core design in UK reactors

    International Nuclear Information System (INIS)

    Davies, M.W.

    1996-01-01

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

  10. Sensing capabilities of graphite based MR elastomers

    International Nuclear Information System (INIS)

    Tian, T F; Li, W H; Deng, Y M

    2011-01-01

    This paper presents both experimental and theoretical investigations of the sensing capabilities of graphite based magnetorheological elastomers (MREs). In this study, eight MRE samples with varying graphite weight fractions were fabricated and their resistance under different magnetic fields and external loadings were measured with a multi-meter. With an increment of graphite weight fraction, the resistance of MRE sample decreases steadily. Higher magnetic fields result in a resistance increase. Based on an ideal assumption of a perfect chain structure, a mathematical model was developed to investigate the relationship between the MRE resistance with external loading. In this model, the current flowing through the chain structure consists of both a tunnel current and a conductivity current, both of which depend on external loadings. The modelling parameters have been identified and reconstructed from comparison with experimental results. The comparison indicates that both experimental results and modelling predictions agree favourably well

  11. Cluster Ion Implantation in Graphite and Diamond

    DEFF Research Database (Denmark)

    Popok, Vladimir

    2014-01-01

    Cluster ion beam technique is a versatile tool which can be used for controllable formation of nanosize objects as well as modification and processing of surfaces and shallow layers on an atomic scale. The current paper present an overview and analysis of data obtained on a few sets of graphite...... and diamond samples implanted by keV-energy size-selected cobalt and argon clusters. One of the emphases is put on pinning of metal clusters on graphite with a possibility of following selective etching of graphene layers. The other topic of concern is related to the development of scaling law for cluster...... implantation. Implantation of cobalt and argon clusters into two different allotropic forms of carbon, namely, graphite and diamond is analysed and compared in order to approach universal theory of cluster stopping in matter....

  12. Graphite based Schottky diodes formed semiconducting substrates

    Science.gov (United States)

    Schumann, Todd; Tongay, Sefaattin; Hebard, Arthur

    2010-03-01

    We demonstrate the formation of semimetal graphite/semiconductor Schottky barriers where the semiconductor is either silicon (Si), gallium arsenide (GaAs) or 4H-silicon carbide (4H-SiC). The fabrication can be as easy as allowing a dab of graphite paint to air dry on any one of the investigated semiconductors. Near room temperature, the forward-bias diode characteristics are well described by thermionic emission, and the extracted barrier heights, which are confirmed by capacitance voltage measurements, roughly follow the Schottky-Mott relation. Since the outermost layer of the graphite electrode is a single graphene sheet, we expect that graphene/semiconductor barriers will manifest similar behavior.

  13. 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.)

  14. Nondestructive evaluation of nuclear-grade graphite

    Science.gov (United States)

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

    2012-05-01

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

  15. Electrostatic Manipulation of Graphene On Graphite

    Science.gov (United States)

    Untiedt, Carlos; Rubio-Verdu, Carmen; Saenz-Arce, Giovanni; Martinez-Asencio, Jesús; Milan, David C.; Moaied, Mohamed; Palacios, Juan J.; Caturla, Maria Jose

    2015-03-01

    Here we report the use of a Scanning Tunneling Microscope (STM) under ambient and vacuum conditions to study the controlled exfoliation of the last layer of a graphite surface when an electrostatic force is applied from a STM tip. In this work we have focused on the study of two parameters: the applied voltage needed to compensate the graphite interlayer attractive force and the one needed to break atomic bonds to produce folded structures. Additionally, we have studied the influence of edge structure in the breaking geometry. Independently of the edge orientation the graphite layer is found to tear through the zig-zag direction and the lifled layer shows a zig-zag folding direction. Molecular Dinamics simulations and DFT calculations have been performed to understand our results, showing a strong correlation with the experiments. Comunidad Valenciana through Prometeo project.

  16. THE EFFECT OF APPLIED STRESS ON THE GRAPHITIZATION OF PYROLYTIC GRAPHITE

    Energy Technology Data Exchange (ETDEWEB)

    Bragg, R H; Crooks, D D; Fenn, Jr, R W; Hammond, M L

    1963-06-15

    Metallographic and x-ray diffraction studies were made of the effect of applied stress at high temperature on the structure of pyrolytic graphite (PG). The dominant factor was whether the PG was above or below its graphitization temperature, which, in turn, was not strongly dependent on applied stress. Below the graphitization temperature, the PG showed a high proportion of disordered layers (0.9), a fairly large mean tilt angle (20 deg ) and a small crystailite size (La --150 A). Fracture occurred at low stress and strain and the materiai exhibited a high apparent Young's modulus ( approximates 4 x 10/sup 6/ psi). Above the graphitization temperature, graphitization was considerably enhanced by strain up to about 8%. The disorder parameter was decreased from a zero strain value of 0.3 to 0.l5 with strain, the mean tilt angle was decreased to 4 deg , and a fivefold increase in crystallite size occurred. When the strainenhanced graphitization was complete, the material exhibited a low apparent modulus ( approximates 0.5 x 10/sup 6/ psi) and large plastic strains (>100%) for a constant stress ( approximates 55 ksi). Graphitization was shown to be a spontaneous process that is promoted by breaking cross-links thermally, and the process is furthered by chemical attack and plastic strain. (auth)

  17. Cementation of Nuclear Graphite Using Geopolymers

    International Nuclear Information System (INIS)

    Girke, N.A.; Steinmetz, H-J.; Bukaemsky, A.; Bosbach, D.; Hermann, E.; Griebel, I.

    2016-01-01

    Geopolymers are solid aluminosilicate materials usually formed by alkali hydroxide or alkali silicate activation of solid precursors such as coal fly ash, calcined clay and/or metallurgical slag. Today the primary application of geopolymer technology is in the development of alternatives to Portland-based cements. Variations in the ratio of aluminium to silicon, and alkali to silicon or addition of structure support, produce geopolymers with different physical and mechanical properties. These materials have an amorphous three-dimensional structure that gives geopolymers certain properties, such as fire and acid resistance, low leach rate, which make them an ideal substitute for ordinary Portland cement (OPC) in a wide range of applications especially in conditioning and storage of radioactive waste. Therefore investigations have been initiated on how and to which amount graphite as a hydrophobic material can be mixed with cement or concrete to form stable waste products and which concretes fulfil the necessary specifications best. As a result, geopolymers have been identified as a promising matrix for graphite containing nuclear wastes. With geopolymers, both favourable properties in the cementation process and a high long time structural stability of the products can be achieved. Investigations include: • direct mixing of graphite with geopolymers with or without sand as a mechanically stabilizing medium; • production of cement-graphite granulates as intermediate products and embedding of these granulates in geopolymer; • coating of formed graphite pieces with geopolymer.The report shows that carbon in the form of graphite can both be integrated with different grain size spectra as well as shaped in the hydraulic binder geopolymer and meets the requirements for a stable long-term immobilisation. (author)

  18. Effects of the Air Flow Rate on The Oxidation of NBG-18 and 25 Nuclear Graphite Grades

    International Nuclear Information System (INIS)

    Chi, Se-Hwan; Kim, Gen-Chan; Jang, Joon-Hee

    2007-01-01

    For a VHTR, graphite oxidation is regarded as a critical phenomenon for degrading the integrity of graphite components under normal or abnormal conditions. The oxidation of a graphite core component can occur by air which may permeate into the primary coolant operation and/or by impurities contained in the He coolant, or by air ingress during a severe accident. It is well known that the oxidation properties of a graphite are highly dependent on the source of raw materials, impurities, microstructures (crystallites, pore structure), and on the processing and environmental parameters, such as the forming methods, the coolant type, moisture and impurity content, temperature, flow rate and the oxygen potential of the coolants. A lot of work has been performed on the oxidation of graphite since the 1960s, and, for example, in the case of the temperature, a widely accepted oxidation model on the effects of a temperature has already been developed. However, in the case of the flow rate, even for its expected effects in a VHTR, for example, as to the expected changes in the bypass flow (10-20 %) during an operation, no systematic works have been performed. In this respect, as a preliminary study, the effects of an air flow rate on the oxidation of NBG-18 and 25 nuclear graphite were investigated

  19. Graphite crystals grown within electromagnetically levitated metallic droplets

    International Nuclear Information System (INIS)

    Amini, Shaahin; Kalaantari, Haamun; Mojgani, Sasan; Abbaschian, Reza

    2012-01-01

    Various graphite morphologies were observed to grow within the electromagnetically levitated nickel–carbon melts, including primary flakes and spheres, curved surface graphite and eutectic flakes, as well as engulfed and entrapped particles. As the supersaturated metallic solutions were cooled within the electromagnetic (EM) levitation coil, the primary graphite flakes and spheres formed and accumulated near the periphery of the droplet due to EM circulation. The primary graphite islands, moreover, nucleated and grew on the droplet surface which eventually formed a macroscopic curved graphite crystal covering the entire liquid. Upon further cooling, the liquid surrounding the primary graphite went under a coupled eutectic reaction while the liquid in the center formed a divorced eutectic due to EM mixing. This brought about the formation of graphite fine flakes and agglomerated particles close to the surface and in the center of the droplet, respectively. The graphite morphologies, growth mechanisms, defects, irregularities and growth instabilities were interpreted with detailed optical and scanning electron microscopies.

  20. Expansion and exfoliation of graphite to form graphene

    KAUST Repository

    Patole, Shashikan P.; Da Costa, Pedro M. F. J.

    2017-01-01

    Graphene production methods are described based on subjecting non- covalent graphite intercalated compounds, such as graphite bisulfate, to expansion conditions such as shocks of heat and/or microwaves followed by turbulence-assisted exfoliation

  1. A 2-D nucleation-growth model of spheroidal graphite

    International Nuclear Information System (INIS)

    Lacaze, Jacques; Bourdie, Jacques; Castro-Román, Manuel Jesus

    2017-01-01

    Analysis of recent experimental investigations, in particular by transmission electron microscopy, suggests spheroidal graphite grows by 2-D nucleation of new graphite layers at the outer surface of the nodules. These layers spread over the surface along the prismatic direction of graphite which is the energetically preferred growth direction of graphite when the apparent growth direction of the nodules is along the basal direction of graphite. 2-D nucleation-growth models first developed for precipitation of pure substances are then adapted to graphite growth from the liquid in spheroidal graphite cast irons. Lateral extension of the new graphite layers is controlled by carbon diffusion in the liquid. This allows describing quantitatively previous experimental results giving strong support to this approach.

  2. Coordinated Isotopic and TEM Studies of Presolar Graphites from Murchison

    Science.gov (United States)

    Croat, T. K.; Stadermann, F. J.; Zinner, E.; Bernatowicz, T. J.

    2004-03-01

    TEM and NanoSIMS investigations of the same presolar Murchison KFC graphites revealed high Zr, Mo, and Ru content in refractory carbides within the graphites. Along with isotopically light carbon, these suggest a low-metallicity AGB source.

  3. Thermal Properties of G-348 Graphite

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-01

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

  4. London forces in highly oriented pyrolytic graphite

    Directory of Open Access Journals (Sweden)

    L.V. Poperenko

    2017-07-01

    Full Text Available Surface of highly oriented pyrolytic graphite with terrace steps was studied using scanning tunneling microscopy with high spatial resolution. Spots with positive and negative charges were found in the vicinity of the steps. Values of the charges depended both on the microscope needle scan velocity and on its motion direction. The observed effect was theoretically explained with account of London forces that arise between the needle tip and the graphite surface. In this scheme, a terrace step works as a nanoscale diode for surface electric currents.

  5. Large Scale Reduction of Graphite Oxide Project

    Science.gov (United States)

    Calle, Carlos; Mackey, Paul; Falker, John; Zeitlin, Nancy

    2015-01-01

    This project seeks to develop an optical method to reduce graphite oxide into graphene efficiently and in larger formats than currently available. Current reduction methods are expensive, time-consuming or restricted to small, limited formats. Graphene has potential uses in ultracapacitors, energy storage, solar cells, flexible and light-weight circuits, touch screens, and chemical sensors. In addition, graphite oxide is a sustainable material that can be produced from any form of carbon, making this method environmentally friendly and adaptable for in-situ reduction.

  6. Chemical atomization of graphite by H+ ions

    International Nuclear Information System (INIS)

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

    A simple model of the mechanism of chemical atomization is given, on whose basis a decrease in chemical atomization is qualitatively predicted for high temperatures. Mass spectrometric investigations of the atomization products cited, which found CH 4 and CH 3 molecules during the irradiation of graphite and H + ions thereby confirmed the presence of chemical atomization. A relationship of S and temperature of graphite T during irradiation was obtained which showed a decrease in the coefficient of atomization of a high temperature. (U.S.)

  7. The electrochemical properties of graphite and carbon

    International Nuclear Information System (INIS)

    Yeager, E.; Gupta, S.; Molla, J.A.

    1983-01-01

    Carbon and graphite are often used as supports for electrocatalysts, but also have an electrocatalytic function in such electrode reactions as O 2 reduction in alkaline electrolytes, Cl 2 generation in brine and SOCl 2 reduction in lithium-thionyl chloride batteries. These catalytic functions involve specific chemical functional groups bound to the carbon and graphite surfaces. The factors controlling O 2 reduction with various types of carbon electrodes of both low and high surface area are reviewed. Of particular importance is the role of hydrogen peroxide. The role of the functionality of the carbon in the electrocatalysis will be discussed

  8. Radiation creep of graphite. An introduction

    Energy Technology Data Exchange (ETDEWEB)

    Blackstone, R [Commission of the European Communities, Petten (Netherlands). Joint Nuclear Research Center

    1977-03-01

    Graphite, a class of materials with many unique and unusual properties, shows a remarkably high creep ductility under irradiation. As this behaviour compensates to some extent some of the more worrying radiation effects, such as dimensional changes and their strong temperature dependence, it is a property of large technological interest. There are various ways of observing and measuring in-pile creep of graphite, varying in degree of sophistication and in cost, in accuracy and in the type of data that is generated. This paper attempts to review briefly the various experimental methods, and the knowledge generated so far. An indication is given of the areas in which further knowledge is wanted.

  9. Radiation creep of graphite. An introduction

    International Nuclear Information System (INIS)

    Blackstone, R.

    1977-01-01

    Graphite, a class of materials with many unique and unusual properties, shows a remarkably high creep ductility under irradiation. As this behavior compensates to some extent some of the more worrying radiation effects, such as dimensional changes and their strong temperature dependence, it is a property of large technological interest. There are various ways of observing and measuring in-pile creep of graphite, varying in degree of sophistication and in cost, in accuracy and in the type of data that is generated. This paper attempts to review briefly the various experimental methods, and the knowledge generated so far. An indication is given of the areas in which further knowledge is wanted

  10. Radiation creep of graphite. An introduction

    International Nuclear Information System (INIS)

    Blackstone, R.

    1977-01-01

    Graphite, a class of materials with many unique and unusual properties, shows a remarkably high creep ductility under irradiation. As this behaviour compensates to some extent some of the more worrying radiation effects, such as dimensional changes and their strong temperature dependence, it is a property of large technological interest. There are various ways of observing and measuring in-pile creep of graphite, varying in degree of sophistication and in cost, in accuracy and in the type of data that is generated. This paper attempts to review briefly the various experimental methods, and the knowledge generated so far. An indication is given of the areas in which further knowledge is wanted. (Auth.)

  11. Electrical properties of Egyptian natural graphite

    International Nuclear Information System (INIS)

    El-Shazly, O.; El-Wahidy, E.F.; Elanany, N.; Saad, N.A.

    1992-06-01

    The electrical properties of Egyptian natural graphite flakes, obtained from the graphite schists of Wadi Bent, Eastern Desert, were measured. The flakes were ground and compressed into pellets. The standard four probe dc method was used to measure the temperature dependence of the electric resistivity from room temperature down to 12 K. The transverse and longitudinal magnetoresistance were measured in the low magnetic field range at temperatures 300 K, 77 K and 12 K. The transverse magnetoresistance data was used to estimate the average mobility, assuming a simple two-band model. (author). 20 refs, 4 figs, 1 tab

  12. Direct reading spectrochemical analysis of nuclear graphite

    International Nuclear Information System (INIS)

    Roca Adell, M.; Becerro Ruiz, E.; Alvarez Gonzalez, F.

    1964-01-01

    A description is given about the application of a direct-reading spectrometer the Quantometer, to the determination of boron. calcium, iron, titanium and vanadium in nuclear grade graphite. for boron the powdered sample is mixed with 1% cupric fluoride and excited in a 10-amperes direct current arc and graphite electrodes with a crater 7 mm wide and 10 mm deep. For the other elements a smaller crater has been used and dilution with a number of matrices has been investigated; the best results are achieved by employing 25% cupric fluoride. The sensitivity limit for boron is 0,15 ppm. (Author) 21 refs

  13. 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)

  14. Ion beam induced surface graphitization of CVD diamond for x-ray beam position monitor applications

    International Nuclear Information System (INIS)

    Liu, Chian; Shu, D.; Kuzay, T.M.; Wen, L.; Melendres, C.A.; Argonne National Lab., IL

    1996-01-01

    The Advanced Photon Source at ANL is a third-generation synchrotron facility that generates powerful x-ray beams on its undulator beamlines. It is important to know the position and angle of the x- ray beam during experiments. Due to very high heat flux levels, several patented x-ray beam position monitors (XBPM) exploiting chemical vapor deposition (CVD) diamond have been developed. These XBPMs have a thin layer of low-atomic-mass metallic coating so that photoemission from the x rays generate a minute but measurable current for position determination. Graphitization of the CVD diamond surface creates a very thin, intrinsic and conducting layer that can stand much higher temperatures and minimal x-ray transmission losses compared to the coated metallic layers. In this paper, a laboratory sputter ion source was used to transform selected surfaces of a CVD diamond substrate into graphite. The effect of 1-5 keV argon ion bombardment on CVD diamond surfaces at various target temperatures from 200 to 500 C was studied using Auger electron spectroscopy and in-situ electrical resistivity measurements. Graphitization after the ion bombardment has been confirmed and optimum conditions for graphitization studied. Raman spectroscopy was used to identify the overall diamond structure in the bulk of CVD diamond substrate after the ion bombardments. It was found that target temperature plays an important role in stability and electrical conductivity of the irradiated CVD diamonds

  15. STS Observations of Landau Levels at Graphite Surfaces

    OpenAIRE

    Matsui, T.; Kambara, H.; Niimi, Y.; Tagami, K.; Tsukada, M.; Fukuyama, Hiroshi

    2004-01-01

    Scanning tunneling spectroscopy measurements were made on surfaces of two different kinds of graphite samples, Kish graphite and highly oriented pyrolytic graphite (HOPG), at very low temperatures and in high magnetic fields. We observed a series of peaks in the tunnel spectra, which grow with increasing field, both at positive and negative bias voltages. These are associated with Landau quantization of the quasi two-dimensional electrons and holes in graphite in magnetic fields perpendicular...

  16. Electronic structure of incident carbon ions on a graphite surface

    International Nuclear Information System (INIS)

    Kiuchi, Masato; Takeuchi, Takae; Yamamoto, Masao.

    1997-01-01

    The electronic structure of an incident carbon ion on a graphite surface is discussed on the basis of ab initio molecular orbital calculations. A carbon cation forms a covalent bond with the graphite, and a carbon nonion is attracted to the graphite surface through van der Waals interaction. A carbon anion has no stable state on a graphite surface. The charge effects of incident ions become clear upon detailed examination of the electronic structure. (author)

  17. Development of a graphite probe calorimeter for absolute clinical dosimetry

    International Nuclear Information System (INIS)

    Renaud, James; Seuntjens, Jan; Sarfehnia, Arman; Marchington, David

    2013-01-01

    The aim of this work is to present the numerical design optimization, construction, and experimental proof of concept of a graphite probe calorimeter (GPC) conceived for dose measurement in the clinical environment (U.S. provisional patent 61/652,540). A finite element method (FEM) based numerical heat transfer study was conducted using a commercial software package to explore the feasibility of the GPC and to optimize the shape, dimensions, and materials used in its design. A functioning prototype was constructed inhouse and used to perform dose to water measurements under a 6 MV photon beam at 400 and 1000 MU/min, in a thermally insulated water phantom. Heat loss correction factors were determined using FEM analysis while the radiation field perturbation and the graphite to water absorbed dose conversion factors were calculated using Monte Carlo simulations. The difference in the average measured dose to water for the 400 and 1000 MU/min runs using the TG-51 protocol and the GPC was 0.2% and 1.2%, respectively. Heat loss correction factors ranged from 1.001 to 1.002, while the product of the perturbation and dose conversion factors was calculated to be 1.130. The combined relative uncertainty was estimated to be 1.4%, with the largest contributors being the specific heat capacity of the graphite (type B, 0.8%) and the reproducibility, defined as the standard deviation of the mean measured dose (type A, 0.6%). By establishing the feasibility of using the GPC as a practical clinical absolute photon dosimeter, this work lays the foundation for further device enhancements, including the development of an isothermal mode of operation and an overall miniaturization, making it potentially suitable for use in small and composite radiation fields. It is anticipated that, through the incorporation of isothermal stabilization provided by temperature controllers, a subpercent overall uncertainty will be achieved.

  18. Systematic Procedure for Generating Operational Policies to Achieve Target Crystal Size Distribution (CSD) in Batch Cooling Crystallization

    DEFF Research Database (Denmark)

    Abdul Samad, Noor Asma Fazli; Singh, Ravendra; Sin, Gürkan

    2011-01-01

    Batch cooling crystallization is one of the important unit operations involving separation of solid-liquid phases. Usually the most common crystal product qualities are directly related to the crystal size distribution (CSD). However the main difficulty in batch crystallization is to obtain a uni...

  19. Effect of thermal annealing on property changes of neutron-irradiated non-graphitized carbon materials and nuclear graphite

    International Nuclear Information System (INIS)

    Matsuo, Hideto

    1991-06-01

    Changes in dimension of non-graphitized carbon materials and nuclear graphite, and the bulk density, electrical resistivity, Young's modulus and thermal expansivity of nuclear graphite were studied after neutron irradiation at 1128-1483 K and the successive thermal annealing up to 2573 K. Carbon materials showed larger and anisotropic dimensional shrinkage than that of nuclear graphite after the irradiation. The irradiation-induced dimensional shrinkage of carbon materials decreased during annealing at temperatures from 1773 to 2023 K, followed by a slight increase at higher temperatures. On the other hand, the irradiated nuclear graphite hardly showed the changes in length, density and thermal expansivity under the thermal annealing, but the electrical resistivity and Young's modulus showed a gradual decrease with annealing temperature. It has been clarified that there exists significant difference in the effect of thermal annealing on irradiation-induced dimensional shrinkage between graphitized nuclear graphite and non-graphitized carbon materials. (author)

  20. Fracture behavior of nuclear graphites under tensile impact loading

    International Nuclear Information System (INIS)

    Ugachi, Hirokazu; Ishiyama, Shintaro; Eto, Motokuni

    1994-01-01

    Impact tensile strength test was performed with two kinds of HTTR graphites, fine grained isotropic graphite, IG-11 and coarse grained near isotropic graphite, PGX and deformation and fracture behavior under the strain rate of over 100s -1 was measured and the following results were derived: (1) Tensile strength for IG-11 graphite does not depend on the strain rate less than 1 s -1 , but over 1 s -1 , tensile strength for IG-11 graphite increase larger than that measured under 1 s -1 . At the strain rate more than 100 s -1 , remarkable decrease of tensile strength for IG-11 graphite was found. Tensile strength of PGX graphite does not depend on the strain rate less than 1 s -1 , but beyond this value, the sharp tensile strength decrease occurs. (2) Under 100 s -1 , fracture strain for both graphites increase with increase of strain rate and over 100 s -1 , drastic increase of fracture strain for IG-11 graphite was found. (3) At the part of gage length, volume of specimen increase with increase of tensile loading level and strain rate. (4) Poisson's ratio for both graphites decrease with increase of tensile loading level and strain rate. (5) Remarkable change of stress-strain curve for both graphites under 100 s -1 was not found, but over 100 s -1 , the slope of these curve for IG-11 graphite decrease drastically. (author)

  1. Oxidation behavior of IG and NBG nuclear graphites

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Woong-Ki; Kim, Byung-Joo [Jeonju Institute of Machinery and Carbon Composites Palbokdong-2ga, 817, Jeonju, Jeollabuk-do 561-844 (Korea, Republic of); Kim, Eung-Seon; Chi, Se-Hwan [Dept. of Nuclear Hydrogen Project, Korea Atomic Energy Research Institute, Daejeon 305-353 (Korea, Republic of); Park, Soo-Jin, E-mail: sjpark@inha.ac.k [Dept. of Chemistry, Inha Univ., 253, Nam-gu, Incheon 402-751 (Korea, Republic of)

    2011-01-15

    Graphical abstract: Water contact angles on nuclear graphite before and after oxidation treatments: the pictures show the contact angles obtained under deionized water on oxidation-treated and untreated nuclear graphite. The water contact angles are decreased after oxidation due to the increase in the hydrophilic. Display Omitted Research highlights: The average pore size of graphites shows an increase after the oxidation treatments. They also show that oxidation produces the surface functional groups on the graphite surfaces. The surface area of each graphite behaves in a unique manner. - Abstract: This work studies the oxidation-induced characteristics of four nuclear graphites (NBG-17, NBG-25, IG-110, and IG-430). The oxidation characteristics of the nuclear graphites were measured at 600 {sup o}C. The surface properties of the oxidation graphites were characterized by means of scanning electron microscopy, X-ray photoelectron spectroscopy, and contact angle methods. The N{sub 2}/77 K adsorption isotherm characteristics, including the specific surface area and micropore volume, were investigated by means of BET and t-plot methods. The experimental results show an increase in the average pore size of graphites; they also show that oxidation produces the surface functional groups on the graphite surfaces. The surface area of each graphite behaves in a unique manner. For example the surface area of NBG-17 increases slightly whereas the surface area of IG-110 increases significantly. This result confirms that the original surface state of each graphite is unique.

  2. Influence of irradiation on high-strength graphites

    International Nuclear Information System (INIS)

    Virgil'ev, Yu.S.; Grebennik, V.N.; Kalyagina, I.P.

    1989-01-01

    To ensure efficiency of the graphite elements of the construction of the masonry of reactors, the graphite must possess high radiation stability, strength, and heat resistance. In this connection, the physical properties of graphites based on uncalcined petroleum coke with a binder - high-temperature hard coal pitch - the amount of which reaches 40% are considered in this paper

  3. Porous graphite electrodes for rechargeable ion-transfer batteries

    Energy Technology Data Exchange (ETDEWEB)

    Novak, P; Scheifele, W; Haas, O [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1997-06-01

    The influence of preparation pressure and pore-forming additives on the properties of graphite-based, Li{sup +}-intercalating electrodes for ion-transfer batteries have been investigated. The electrochemical performance of graphite electrodes could be improved by adjusting the porosity. Specific charge of >300 Ah/kg (with respect to the graphite mass) could be achieved. (author) 4 figs., 2 refs.

  4. AA antiproton production target

    CERN Multimedia

    CERN PhotoLab

    1979-01-01

    The first version of the antiproton production target was a tungsten rod, 11 cm long (actually a row of 11 rods, each 1 cm long) and 3 mm in diameter. The rod was embedded in graphite, pressure-seated into an outer casing made of stainless steel. The casing had fins for forced-air cooling. In this picture, the 26 GeV high-intensity beam from the PS enters from the right, where a scintillator screen, with circles every 5 mm in radius, permits precise aim at the target centre. See also 7903034 and 7905094.

  5. Preparation of graphite derivatives by selective reduction of graphite oxide and isocyanate functionalization

    Czech Academy of Sciences Publication Activity Database

    Kumar, A. R. S. S.; Piana, Francesco; Mičušík, M.; Pionteck, J.; Banerjee, S.; Voit, B.

    2016-01-01

    Roč. 182, 1 October (2016), s. 237-245 ISSN 0254-0584 Institutional support: RVO:61389013 Keywords : graphite oxide * surface modification * conductive nanoparticles Subject RIV: CD - Macromolecular Chemistry Impact factor: 2.084, year: 2016

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

    Science.gov (United States)

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

    2017-10-01

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

  7. A pupal transcriptomic screen identifies Ral as a target of store-operated calcium entry in Drosophila neurons

    OpenAIRE

    Richhariya, Shlesha; Jayakumar, Siddharth; Abruzzi, Katharine; Rosbash, Michael; Hasan, Gaiti

    2017-01-01

    Transcriptional regulation by Store-operated Calcium Entry (SOCE) is well studied in non-excitable cells. However, the role of SOCE has been poorly documented in neuronal cells with more complicated calcium dynamics. Previous reports demonstrated a requirement for SOCE in neurons that regulate Drosophila flight bouts. We refine this requirement temporally to the early pupal stage and use RNA-sequencing to identify SOCE mediated gene expression changes in the developing Drosophila pupal nervou...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-15

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

  9. Polyphase diffusion of fission products in graphite

    International Nuclear Information System (INIS)

    Dannert, V.

    1989-05-01

    The report attempts to give an introduction into the subject of fission product transport in nuclear graphite and results in an extended proposal of a transport-model. Beginning with a rough description of the graphite in question, an idea about the physical transport-phenomena in graphite is developed. Some of the basic experimental methods, especially techniques of porosimetry, determination of sorption-isotherms and of course several transport-experiments, are briefly described and their results are discussed. Some of the most frequent transport models are introduced and assessed with the criteria emphasized in this report. An extended model is proposed including the following main ideas: The transport of the fission-products is regarded as a two-phase-diffusion process through the open pores of the graphite. The two phases are: surface-diffusion and gas-diffusion. A time-dependent coupling of the two diffusion-phases by sorption-isotherms and a concentration-dependence of the surface diffusion coefficient, also related to the physical behaviour of the sorption-isotherms, are the basic properties of the proposed model. (orig./HP) [de

  10. US graphite reactor D ampersand D experience

    International Nuclear Information System (INIS)

    Garrett, S.M.K.; Williams, N.C.

    1997-02-01

    This report describes the results of the U.S. Graphite Reactor Experience Task for the Decommissioning Strategy Plan for the Leningrad Nuclear Power Plant (NPP) Unit 1 Study. The work described in this report was performed by the Pacific Northwest National Laboratory (PNNL) for the Department of Energy (DOE)

  11. Analysis of a T-10 graphite limiter

    International Nuclear Information System (INIS)

    Hildebrandt, D.; Laux, M.; Lingertat, J.; Pech, P.; Reiner, H.D.; Strusny, H.; Wolff, H.

    1981-01-01

    Parts of a T-10 graphite limiter used during ohmic heated discharges have been investigated. Erosion and deposition phenomena have been studied by morphological and elemental surface analysis methods. From the results estimates of the plasma parameters near the limiter surface have been made. (orig.)

  12. Raw materials for reflector graphite (for reactors)

    International Nuclear Information System (INIS)

    Wilhelmi, G.; Mindermann, D.

    1992-01-01

    The manufacturing concept for the core components of German high temperature reactor (HTR) types of graphite was previously entirely directed to the use of German tar coke (St coke). As the plants for producing this material no longer complied technically with the current environmental protection requirements, one had to assume that they would soon be shut down. To prevent bottlenecks in the erection of future HTR plants, alternative cokes produced by modern processes by Japanese manufacturers were checked for their suitability for the manufacture of reactor graphite. This report describes the investigations carried out on these materials from the safe delayed coking process. The project work, apart from analysis of the main data of the candidate coke considered, included the processing of the raw materials into directly and secondarily extruded graphite rods on the laboratory scale, including characterisation. As the results show, the material data achieved with the previous raw material can be reproduced with Japanese St coke. The tar coke LPC-A from the Nippon Steel Chemical Co., Ltd was decided on as the new standard coke for manufacturing reflector graphite. (orig.) With 15 tabs., 2 figs [de

  13. Formation of dislocation dipoles in irradiated graphite

    International Nuclear Information System (INIS)

    Niwase, Keisuke

    2005-01-01

    Recently, we have proposed a dislocation dipole accumulation model to explain the irradiation-induced amorphization of graphite. However, the structure of dislocation dipole in the hexagonal networks is still an open question at the atomic-level. In this paper, we propose a possible formation process of the dislocation dipole

  14. Functional interface of polymer modified graphite anode

    Energy Technology Data Exchange (ETDEWEB)

    Komaba, S.; Ozeki, T.; Okushi, K. [Department of Applied Chemistry, Tokyo University of Science, 1-3 Kagurazaka, Shinjuku, Tokyo 162-8601 (Japan)

    2009-04-01

    Graphite electrodes were modified by polyacrylic acid (PAA), polymethacrylic acid (PMA), and polyvinyl alcohol (PVA). Their electrochemical properties were examined in 1 mol dm{sup -3} LiClO{sub 4} ethylene carbonate:dimethyl carbonate (EC:DMC) and propylene carbonate (PC) solutions as an anode of lithium ion batteries. Generally, lithium ions hardly intercalate into graphite in the PC electrolyte due to a decomposition of the PC electrolyte at ca. 0.8 V vs. Li/Li{sup +}, and it results in the exfoliation of the graphene layers. However, the modified graphite electrodes with PAA, PMA, and PVA demonstrated the stable charge-discharge performance due to the reversible lithium intercalation not only in the EC:DMC but also in the PC electrolytes since the electrolyte decomposition and co-intercalation of solvent were successfully suppressed by the polymer modification. It is thought that these improvements were attributed to the interfacial function of the polymer layer on the graphite which interacted with the solvated lithium ions at the electrode interface. (author)

  15. Functional interface of polymer modified graphite anode

    Science.gov (United States)

    Komaba, S.; Ozeki, T.; Okushi, K.

    Graphite electrodes were modified by polyacrylic acid (PAA), polymethacrylic acid (PMA), and polyvinyl alcohol (PVA). Their electrochemical properties were examined in 1 mol dm -3 LiClO 4 ethylene carbonate:dimethyl carbonate (EC:DMC) and propylene carbonate (PC) solutions as an anode of lithium ion batteries. Generally, lithium ions hardly intercalate into graphite in the PC electrolyte due to a decomposition of the PC electrolyte at ca. 0.8 V vs. Li/Li +, and it results in the exfoliation of the graphene layers. However, the modified graphite electrodes with PAA, PMA, and PVA demonstrated the stable charge-discharge performance due to the reversible lithium intercalation not only in the EC:DMC but also in the PC electrolytes since the electrolyte decomposition and co-intercalation of solvent were successfully suppressed by the polymer modification. It is thought that these improvements were attributed to the interfacial function of the polymer layer on the graphite which interacted with the solvated lithium ions at the electrode interface.

  16. Removal of iron from impure graphites

    International Nuclear Information System (INIS)

    Growcock, F.B.; Heiser, J.

    1979-01-01

    Iron-impregnated and ash-rich graphites have been purified by leaching with gaseous I 2 at 900 0 C. With addition of H 2 , the rate of removal of impurity iron can be markedly increased and becomes comparable to that obtained with Cl 2 . I 2 has an advantage in that it can also volatilize Ca and perhaps Ba and Sr

  17. 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.)

  18. Thermoexpanded graphite modification by titanium dioxide

    International Nuclear Information System (INIS)

    Semko, L.S.; Gorbik, P.P.; Chujko, O.O.; Kruchek, Ya.Yi.; Dzyubenko, L.S.; Orans'ka, O.Yi.

    2006-01-01

    A method of the synthesis of thermoexpanded graphite (TEG) powders coated by titanium dioxide is developed. The conversion of n-buthylorthotitanate into TiO 2 on the TEG surface is investigated. The optimal parameters of the synthesis and the structure of titanium dioxide clusters on the TEG surface are determined

  19. GRAPHITIZED STEELS IN MACHINE-BUILDING

    Directory of Open Access Journals (Sweden)

    I. V. Akimov

    2010-01-01

    Full Text Available It is shown that graphitized steels in some cases due to its intermediate disposition by structure and characteristics among low-carbon steels and cast irons, can provide the necessary combination of characteristics of construction material and consequently to increase safety and durability of details of metallurgical and machinebuilding industry machines.

  20. The uncertain future for nuclear graphite disposal: Crisis or opportunity?

    International Nuclear Information System (INIS)

    Wickham, A.J.; Neighbour, G.B.; Dubourg, M.

    2001-01-01

    to be retained in interim surface storage. It appears to the authors to be an appropriate time to review the situation closely and to consider the problem entirely on the basis of established scientific fact and a comparative risk analysis, unfettered by the perceived political constraints to which the nuclear operators and decommissioning organisations are necessarily subjected. This paper seeks to raise again for scientific debate all of these possibilities for graphite disposal, without prejudice necessarily to political constraints, in order to try to focus on genuinely the most satisfactory outcome to all concerned. (author)

  1. Metal/graphite-composite materials for fusion device

    International Nuclear Information System (INIS)

    Kneringer, G.; Kny, E.; Fischer, W.; Reheis, N.; Staffler, R.; Samm, U.; Winter, J.

    1995-01-01

    The utilization of graphite as a structural material depends to an important extent on the availability of a joining technique suitable for the production of reliable large scale metal/graphite-composites. This study has been conducted to evaluate vacuum brazes and procedures for graphite and metals which can be used in fusion applications up to about 1500 degree C. The braze materials included: AgCuTi, CuTi, NiTi, Ti, ZrTi, Zr. Brazing temperatures ranged from 850 degree C to 1900 degree C. The influence of graphite quality on wettability and pore-penetration of the braze has been investigated. Screening tests of metal/graphite-assemblies with joint areas exceeding some square-centimeters have shown that they can only successfully be produced when graphite is brazed to a metal, such as tungsten or molybdenum with a coefficient of thermal expansion closely matching that of graphite. Therefore all experimental work on evaluation of joints has been concentrated on molybdenum/graphite brazings. The tensile strength of molybdenum/graphite-composites compares favorably with the tensile strength of bulk graphite from room temperature close to the melting temperature of the braze. In electron beam testing the threshold damage line for molybdenum/graphite-composites has been evaluated. Results show that even composites with the low melting AgCuTi-braze are expected to withstand 10 MW/m 2 power density for at least 10 3 cycles. Limiter testing in TEXTOR shows that molybdenum/graphite-segments with 3 mm graphite brazed on molybdenum-substrate withstand severe repeated TEXTOR plasma discharge conditions without serious damage. Results prove that actively cooled components on the basis of a molybdenum/graphite-composite can sustain a higher heat flux than bulk graphite alone. (author)

  2. 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

  3. 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).

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  5. Conception and operation of a 10 kCi liquid tritium target for the study of tritium nucleus by electron diffusion: 3H(e,e)

    International Nuclear Information System (INIS)

    Juster, F.P.

    1986-06-01

    We describe the conception and operation of an experimental setup, specially suited for the study of the nuclear structure of tritium by elastic electron scattering at intermediate energy. The experiment has been conducted at the ALS 700 MeV electron linac (Saclay, France). The radioactive nature of tritium has led to the design of a new target, suited for handling reliably ten kilocuries of tritium (one gram). The tritium is contained in three sealed envelopes. Initially a high pressure gas (23 bars) at room temperature, the tritium is cooled down to liquefaction by thermal conduction through a solid, without breaking any seal. The beam path and the scattered trajectories cross thin metallic windows. Additional protection, during the presence of personnel, is provided by a heavy container, remotely operated. Any leak in the containement vessels is detected by changes in pressure and/or temperature gauges, monitored by two independent processors. These processors handle the operation of the outer container, the beam switching and the spare venting system. No tritium leak has been detected during a total six-week run. The tritium liquefies in a cylindrical target, 5 cm long and 1 cm in diameter. A beam of 10 microamperes, in the 200-700 MeV, has been measured. The charge and magnetic form factors of tritium have been measured up to a momentum transfer of 31.3 fm -2 [fr

  6. Graphite Oxidation Simulation in HTR Accident Conditions

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed

    2012-10-19

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

  7. Graphitization in Carbon MEMS and Carbon NEMS

    Science.gov (United States)

    Sharma, Swati

    Carbon MEMS (CMEMS) and Carbon NEMS (CNEMS) are an emerging class of miniaturized devices. Due to the numerous advantages such as scalable manufacturing processes, inexpensive and readily available precursor polymer materials, tunable surface properties and biocompatibility, carbon has become a preferred material for a wide variety of future sensing applications. Single suspended carbon nanowires (CNWs) integrated on CMEMS structures fabricated by electrospinning of SU8 photoresist on photolithographially patterned SU8 followed by pyrolysis are utilized for understanding the graphitization process in micro and nano carbon materials. These monolithic CNW-CMEMS structures enable the fabrication of very high aspect ratio CNWs of predefined length. The CNWs thus fabricated display core---shell structures having a graphitic shell with a glassy carbon core. The electrical conductivity of these CNWs is increased by about 100% compared to glassy carbon as a result of enhanced graphitization. We explore various tunable fabrication and pyrolysis parameters to improve graphitization in the resulting CNWs. We also suggest gas-sensing application of the thus fabricated single suspended CNW-CMEMS devices by using the CNW as a nano-hotplate for local chemical vapor deposition. In this thesis we also report on results from an optimization study of SU8 photoresist derived carbon electrodes. These electrodes were applied to the simultaneous detection of traces of Cd(II) and Pb(II) through anodic stripping voltammetry and detection limits as low as 0.7 and 0.8 microgL-1 were achieved. To further improve upon the electrochemical behavior of the carbon electrodes we elucidate a modified pyrolysis technique featuring an ultra-fast temperature ramp for obtaining bubbled porous carbon from lithographically patterned SU8. We conclude this dissertation by suggesting the possible future works on enhancing graphitization as well as on electrochemical applications

  8. Validation of electro-thermal simulation with experimental data to prepare online operation of a molten salt target at ISOLDE for the Beta Beams

    CERN Document Server

    Cimmino, S; Marzari, S; Stora, T

    2013-01-01

    The main objective of the Beta Beams is to study oscillation property of pure electrons neutrinos. It produces high energy beams of pure electron neutrinos and anti-neutrinos for oscillation experiments by beta decay of He-6 and Ne-18 radioactive ion beams, stored in a decay ring at gamma = 100. The production of He-6 beam has already been accomplished using a thick beryllium oxide target. However, the production of the needed rate of Ne-18 has proven to be more challenging. In order to achieve the requested yield for Ne-18 a new high power target design based on a circulating molten salt loop has been proposed. To verify some elements of the design, a static molten salt target prototype has been developed at ISOLDE and operated successfully. This paper describes the electro-thermal study of the molten salt target taking into account the heat produced by Joule effect, radiative heat exchange, active water cooling due to forced convection and air passive cooling due to natural convection. The numerical results...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-15

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

  10. Ground-based characterization of Hayabusa2 mission target asteroid 162173 Ryugu: constraining mineralogical composition in preparation for spacecraft operations

    Science.gov (United States)

    Le Corre, Lucille; Sanchez, Juan A.; Reddy, Vishnu; Takir, Driss; Cloutis, Edward A.; Thirouin, Audrey; Becker, Kris J.; Li, Jian-Yang; Sugita, Seiji; Tatsumi, Eri

    2018-03-01

    Asteroids that are targets of spacecraft missions are interesting because they present us with an opportunity to validate ground-based spectral observations. One such object is near-Earth asteroid (NEA) (162173) Ryugu, which is the target of the Japanese Space Agency's (JAXA) Hayabusa2 sample return mission. We observed Ryugu using the 3-m NASA Infrared Telescope Facility on Mauna Kea, Hawaii, on 2016 July 13 to constrain the object's surface composition, meteorite analogues, and link to other asteroids in the main belt and NEA populations. We also modelled its photometric properties using archival data. Using the Lommel-Seeliger model we computed the predicted flux for Ryugu at a wide range of viewing geometries as well as albedo quantities such as geometric albedo, phase integral, and spherical Bond albedo. Our computed albedo quantities are consistent with results from Ishiguro et al. Our spectral analysis has found a near-perfect match between our spectrum of Ryugu and those of NEA (85275) 1994 LY and Mars-crossing asteroid (316720) 1998 BE7, suggesting that their surface regoliths have similar composition. We compared Ryugu's spectrum with that of main belt asteroid (302) Clarissa, the largest asteroid in the Clarissa asteroid family, suggested as a possible source of Ryugu by Campins et al. We found that the spectrum of Clarissa shows significant differences with our spectrum of Ryugu, but it is similar to the spectrum obtained by Moskovitz et al. The best possible meteorite analogues for our spectrum of Ryugu are two CM2 carbonaceous chondrites, Mighei and ALH83100.

  11. Grafting polyethylenimine with quinoline derivatives for targeted imaging of intracellular Zn{sup 2+} and logic gate operations

    Energy Technology Data Exchange (ETDEWEB)

    Pan, Yi; Shi, Yupeng; Chen, Junying; Wong, Chap-Mo; Zhang, Heng [Key Laboratory of Sensing Technology and Biomedical Instruments (Guangdong Province), School of Engineering, Sun Yat-Sen University, Guangzhou (China); Li, Mei-Jin [Key Laboratory of Analysis and Detection Technology for Food Safety, Ministry of Education and Fujian Province, Department of Chemistry, Fuzhou University, Fuzhou (China); Li, Cheuk-Wing [Institute of Chinese Medical Sciences, University of Macau (China); Yi, Changqing, E-mail: yichq@mail.sysu.edu.cn [Key Laboratory of Sensing Technology and Biomedical Instruments (Guangdong Province), School of Engineering, Sun Yat-Sen University, Guangzhou (China); Research Institute of Sun Yat-Sen University in Shenzhen, Shenzhen (China)

    2016-12-01

    In this study, a highly sensitive and selective fluorescent Zn{sup 2+} probe which exhibited excellent biocompatibility, water solubility, and cell-membrane permeability, was facilely synthesized in a single step by grafting polyethyleneimine (PEI) with quinoline derivatives. The primary amino groups in the branched PEI can increase water solubility and cell permeability of the probe PEIQ, while quinoline derivatives can specifically recognize Zn{sup 2+} and reduce the potential cytotoxicity of PEI. Basing on fluorescence off-on mechanism, PEIQ demonstrated excellent sensing capability towards Zn{sup 2+} in absolute aqueous solution, where a high sensitivity with a detection limit as low as 38.1 nM, and a high selectivity over competing metal ions and potential interfering amino acids, were achieved. Inspired by these results, elementary logic operations (YES, NOT and INHIBIT) have been constructed by employing PEIQ as the gate while Zn{sup 2+} and EDTA as chemical inputs. Together with the low cytotoxicity and good cell-permeability, the practical application of PEIQ in living cell imaging was satisfactorily demonstrated, emphasizing its wide application in fundamental biology research. - Graphical abstract: The fluorescent Zn{sup 2+} probe, PEIQ, is facilely synthesized by grafting PEI with 8-CAAQ, and demonstrated for the pratical applications in Zn{sup 2+} imaging and implementation of molecular logic operations within biological cells. - Highlights: • PEIQ, fluorescent Zn{sup 2+} probe, is synthesized by grafting PEI with quinoline derivatives. • PEIQ exhibits high sensitivity and selectivity in absolute aqueous solution. • PEIQ is biocompatible, water soluble, and cell-membrane permeable. • Elementary logic operations have been demonstrated for PEIQ/Zn{sup 2+}/EDTA system. • The practical application of PEIQ in living cell imaging is demonstrated.

  12. A study of the coefficient of thermal expansion of nuclear graphites

    International Nuclear Information System (INIS)

    Hacker, P.J.

    2001-02-01

    This thesis presents the results of a study of the Coefficient of Thermal Expansion (CTE) of two grades of nuclear graphite that are used as the moderator in the Magnox and Advanced Gas-Cooled reactors operated in the UK. This work has two main aims, the first is to characterise those elements of the graphite microstructure that control CTE within these materials and to relate these to the effects induced within the reactor. The second is to develop a microstructural model, of general applicability, that can initially be applied to model the CTE changes within the graphites under reactor conditions (neutron irradiation and radiolytic oxidation). These aims have been met by study in three interlinked areas, theoretical, experimental and modelling. Previous to this study, a loose assembly of single crystals together with changes in small scale nanometric porosity (Mrozowski cracks) were used to describe CTE behaviour of nuclear graphite both as-received and under reactor conditions. Within the experimental part of this thesis the graphite nanostructure was studied using, primarily, Transmission Electron Microscopy (TEM). This work concluded that structure on this scale was complex and that the loose assembly of single crystals was a poor microstructural approximation for modelling the CTE of these materials. Other experimental programmes measured the CTE of highly oxidised samples and simulated the effects of irradiation. The former discovered that CTE remained largely unaffected to high weight losses. This insensitivity was explained by ''The Continuous Network Hypothesis'' that was also related to classical percolation theory. The final part of the thesis modelled an abstraction of the key microstructural features identified in the previous parts of the thesis. This approach has been applied to AGR moderator graphite where it has successfully modelled the thermal expansion behaviour of the as-received, irradiated and oxidised material. (author)

  13. Graphite-based detectors of alkali metals for nuclear power plants

    International Nuclear Information System (INIS)

    Kalandarishvili, A.G.; Kuchukhidze, V.A.; Sordiya, T.D.; Shartava, Sh.Sh.; Stepennov, B.S.

    1993-01-01

    The coolants most commonly used in today's fast reactors are alkali metals or their alloys. A major problem in nuclear plant design is leakproofing of the liquid-metal cooling system, and many leak detection methods and safety specifications have been developed as a result. Whatever the safety standards adopted for nuclear plants in different countries, they all rely on the basic fact that control of the contamination and radiation hazards involved requires reliable monitoring equipment. Results are presented of trials with some leak detectors for the alkali-metal circuits of nuclear reactors. The principal component affecting the detector performance is the sensing element. In the detectors graphite was employed, whose laminar structure enables it to absorb efficiently alkali-metal vapors at high temperatures (320--500 K). This produces a continuous series of alkali-metal-graphite solid solutions with distinct electrical, thermal, and other physical properties. The principle of operation of the detectors resides in the characteristic reactions of the metal-graphite system. One detector type uses the change of electrical conductivity of the graphite-film sensor when it is exposed to alkali-metal vapor. In order to minimize the effect of temperature on the resistance the authors prepared composite layers of graphite intercalated with a donor impurity (cesium or barium), and a graphite-nickel material. The addition of a small percentage of cesium, barium, or nickel produces a material whose temperature coefficient of resistance is nearly zero. Used as a sensing element, such a material can eliminate the need for thermostatic control of the detector

  14. Baking and helium glow discharge cleaning of SST-1 tokamak with graphite plasma facing components

    International Nuclear Information System (INIS)

    Semwal, Pratibha; Khan, Ziauddin; Raval, Dilip

    2015-01-01

    Graphite plasma facing components (PFCs) were installed inside SST-1 vacuum vessel. Prior to installation, all the graphite tiles were baked at 1000 °C in a vacuum furnace operated below 1.0 X 10 -5 mbar. However due to the porous structure of graphite, they absorb a significant amount of water vapour from air during the installation process. Rapid desorption of water vapour requires high temperature bake-out of the PFCs at ≥ 250 °C. In SST-1 the PFCs were baked at 250 °C using hot nitrogen gas facility to remove the absorbed water vapour. Also device with large graphite surface area has the disadvantage that a large quantity of hydrogen gets trapped inside it during plasma discharges which makes density control difficult. Helium (He) glow discharge cleaning (GDC) effectively removes this stored hydrogen as well as other impurities like oxygen and hydrocarbon within few nanometers from the surface by particle induced desorption. Before plasma operation in SST-1 tokamak, both baking of PFCs and He-GDC were carried out so that these impurities were removed effectively. The mean desorption yield of hydrogen was found to be 0.48. In this paper, the results of effect of baking and He-GDC experiments of SST-1 will be presented in detail. (author)

  15. Baking and helium glow discharge cleaning of SST-1 Tokamak with graphite plasma facing components

    International Nuclear Information System (INIS)

    Semwal, P; Khan, Z; Raval, D C; Dhanani, K R; George, S; Paravastu, Y; Prakash, A; Thankey, P; Ramesh, G; Khan, M S; Saikia, P; Pradhan, S

    2017-01-01

    Graphite plasma facing components (PFCs) were installed inside the SST-1 vacuum vessel. Prior to installation, all the graphite tiles were baked at 1000 °C in a vacuum furnace operated below 1.0 × 10 -5 mbar. However due to the porous structure of graphite, they absorb a significant amount of water vapour from air during the installation process. Rapid desorption of this water vapour requires high temperature bake-out of the PFCs at ≥ 250 °C. In SST-1 the PFCs were baked at 250 °C using hot nitrogen gas facility to remove the absorbed water vapour. Also device with large graphite surface area has the disadvantage that a large quantity of hydrogen gets trapped inside it during plasma discharges which makes density control difficult. Helium glow discharge cleaning (He-GDC) effectively removes this stored hydrogen as well as other impurities like oxygen and hydrocarbon within few nano-meters from the surface by particle induced desorption. Before plasma operation in SST-1 tokamak, both baking of PFCs and He-GDC were carried out so that these impurities were removed effectively. The mean desorption yield of hydrogen was found to be 0.24. In this paper the results of baking and He-GDC experiments of SST-1 will be presented in detail. (paper)

  16. Baking and helium glow discharge cleaning of SST-1 Tokamak with graphite plasma facing components

    Science.gov (United States)

    Semwal, P.; Khan, Z.; Raval, D. C.; Dhanani, K. R.; George, S.; Paravastu, Y.; Prakash, A.; Thankey, P.; Ramesh, G.; Khan, M. S.; Saikia, P.; Pradhan, S.

    2017-04-01

    Graphite plasma facing components (PFCs) were installed inside the SST-1 vacuum vessel. Prior to installation, all the graphite tiles were baked at 1000 °C in a vacuum furnace operated below 1.0 × 10-5 mbar. However due to the porous structure of graphite, they absorb a significant amount of water vapour from air during the installation process. Rapid desorption of this water vapour requires high temperature bake-out of the PFCs at ≥ 250 °C. In SST-1 the PFCs were baked at 250 °C using hot nitrogen gas facility to remove the absorbed water vapour. Also device with large graphite surface area has the disadvantage that a large quantity of hydrogen gets trapped inside it during plasma discharges which makes density control difficult. Helium glow discharge cleaning (He-GDC) effectively removes this stored hydrogen as well as other impurities like oxygen and hydrocarbon within few nano-meters from the surface by particle induced desorption. Before plasma operation in SST-1 tokamak, both baking of PFCs and He-GDC were carried out so that these impurities were removed effectively. The mean desorption yield of hydrogen was found to be 0.24. In this paper the results of baking and He-GDC experiments of SST-1 will be presented in detail.

  17. New Insights of Graphite Anode Stability in Rechargeable Batteries: Li-Ion Coordination Structures Prevail over Solid Electrolyte Interphases

    KAUST Repository

    Ming, Jun

    2018-01-04

    Graphite anodes are not stable in most noncarbonate solvents (e.g., ether, sulfoxide, sulfone) upon Li ion intercalation, known as an urgent issue in present Li ions and next-generation Li–S and Li–O2 batteries for storage of Li ions within the anode for safety features. The solid electrolyte interphase (SEI) is commonly believed to be decisive for stabilizing the graphite anode. However, here we find that the solvation structure of the Li ions, determined by the electrolyte composition including lithium salts, solvents, and additives, plays a more dominant role than SEI in graphite anode stability. The Li ion intercalation desired for battery operation competes with the undesired Li+–solvent co-insertion, leading to graphite exfoliation. The increase in organic lithium salt LiN(SO2CF3)2 concentration or, more effectively, the addition of LiNO3 lowers the interaction strength between Li+ and solvents, suppressing the graphite exfoliation caused by Li+–solvent co-insertion. Our findings refresh the knowledge of the well-known SEI for graphite stability in metal ion batteries and also provide new guidelines for electrolyte systems to achieve reliable and safe Li–S full batteries.

  18. 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.

  19. Treatment of graphite felt by modified Hummers method for the positive electrode of vanadium redox flow battery

    International Nuclear Information System (INIS)

    Wu, Xiaoxin; Xu, Hongfeng; Shen, Yang; Xu, Pengcheng; Lu, Lu; Fu, Jie; Zhao, Hong

    2014-01-01

    A novel and highly effective treatment based on modified Hummers method was firstly used to improve the electrochemical activity of graphite felt as the positive electrode in all-vanadium redox flow battery (VRFB). The graphite felt was treated by the modified Hummers method and characterized by Fourier transform infrared spectroscopy and scanning electron microscopy. The electrochemical performance of the prepared electrode was evaluated through cyclic voltammetry and electrochemical impedance spectroscopy. Results show that graphite felt treated by modified Hummers method exhibits excellent electrocatalytic activity and reaction rate to vanadium redox couples. In our research, the hydrogen electrode and H 2 replaced the graphite felt and V 2+ /V 3+ couple in the negative side in the VRFB performance test. The coulombic, voltage, and energy efficiencies of the VRFB with the as-prepared electrodes at 50 mA cm −2 are 95.0%, 81.3%, and 77.2%, respectively. These values are much higher than those of the cell-assembled graphite felt electrodes that were conventionally and thermally treated. The graphite felt treated by the modified Hummers method carries more hydrophilic groups, such as–OH, on its defects, which is advantageous in facilitating the redox reaction of vanadium ions, thereby improving the operation efficiency of the vanadium redox flow battery

  20. Pulsed Laser Deposition of Tungsten Thin Films on Graphite

    International Nuclear Information System (INIS)

    Kassem, W.; Tabbal, M.; Roumie, M.

    2011-01-01

    Thin coatings of Tungsten were deposited on substrates fabricated by pre-depositing graphite thin layers on Si(100) wafers. We ablate pure W target using a 20 ns KrF excimer laser (248 nm) in an Ar ambient. The effect of background gas pressure, substrate temperature, and laser fluence, on the properties of the deposited W layers is studied using several techniques including X-Ray Diffraction, Atomic Force Microscopy, surface profilometry, and Rutherford Back-Scattering spectrometry. Our results indicate that the deposited layers consist of the well-crystallized body-centered-cubic α-W phase with bulk-like properties, particularly for films deposited at a substrate temperature of 450 0 C, laser fluence greater than 400mJ, and pressure of about 10mTorr. (author)

  1. 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.)

  2. Ion irradiated graphite exposed to fusion-relevant deuterium plasma

    International Nuclear Information System (INIS)

    Deslandes, Alec; Guenette, Mathew C.; Corr, Cormac S.; Karatchevtseva, Inna; Thomsen, Lars; Ionescu, Mihail; Lumpkin, Gregory R.; Riley, Daniel P.

    2014-01-01

    Graphite samples were irradiated with 5 MeV carbon ions to simulate the damage caused by collision cascades from neutron irradiation in a fusion environment. The ion irradiated graphite samples were then exposed to a deuterium plasma in the linear plasma device, MAGPIE, for a total ion fluence of ∼1 × 10 24 ions m −2 . Raman and near edge X-ray absorption fine structure (NEXAFS) spectroscopy were used to characterize modifications to the graphitic structure. Ion irradiation was observed to decrease the graphitic content and induce disorder in the graphite. Subsequent plasma exposure decreased the graphitic content further. Structural and surface chemistry changes were observed to be greatest for the sample irradiated with the greatest fluence of MeV ions. D retention was measured using elastic recoil detection analysis and showed that ion irradiation increased the amount of retained deuterium in graphite by a factor of four

  3. GRAPHITIZATION OF METASEDIMENTARY ROCKS IN THE WESTERN KONYA

    Directory of Open Access Journals (Sweden)

    Hüseyin KURT

    2000-01-01

    Full Text Available The Paleozoic-Mesozoic metasedimentary rocks in the study area are metacarbonate, metachert, metapelite, metasandstone and metaconglomerate. Graphite layers are 1cm to 2m thick, extend laterally for tens of meters and are intercalated with metasedimentary rocks. Generally, the graphite is black in color, with a well developed cleavage which is concordant with the cleavage of the host rocks. In addition, the crystal and flake graphites formed in metasedimentary rocks are mostly aligned parallel to the cleavage planes. These metamorphic rocks are subjected to shearing and granulation providing structural control for the development of graphite. It was probably this phenomenon that first led to emphasize the relationship between graphite and metasedimentary rocks. Graphite mineralization has been controlled by bedding, microfractures and granulations. Briefly, the metamorphism has converted carbonaceous matter into graphite .

  4. Hydrophilization of graphite using plasma above/in a solution

    Science.gov (United States)

    Hoshino, Shuhei; Kawahara, Kazuma; Takeuchi, Nozomi

    2018-01-01

    A hydrophilization method for graphite is required for applications such as conductive ink. In typical chemical oxidation methods for graphite have the problems of producing many defects in graphite and a large environmental impact. In recent years, the plasma treatment has attracted attention because of the high quality of the treated samples and the low environmental impact. In this study, we proposed an above-solution plasma treatment with a high contact probability of graphite and plasma since graphite accumulates on the solution surface due to its hydrophobicity, which we compared with a so-called solution plasma treatment. Graphite was hydrophilized via reactions with OH radicals generated by the plasma. It was confirmed that hydroxyl and carboxyl groups were modified to the graphite and the dispersibility was improved. The above-solution plasma achieved more energy-efficient hydrophilization than the solution plasma and it was possible to enhance the dispersibility by increasing the plasma-solution contact area.

  5. Crystallization degree change of expanded graphite by milling and annealing

    International Nuclear Information System (INIS)

    Tang Qunwei; Wu Jihuai; Sun Hui; Fang Shijun

    2009-01-01

    Expanded graphite was ball milled with a planetary mill in air atmosphere, and subsequently thermal annealed. The samples were characterized by using X-ray diffraction spectroscopy (XRD), scanning electron microscopy (SEM) and thermal gravimetric analysis (TGA). It was found that in the milling initial stage (less than 12 h), the crystallization degree of the expanded graphite declined gradually, but after milling more than 16 h, a recrystallization of the expanded graphite toke place, and ordered nanoscale expanded graphite was formed gradually. In the annealing initial stage, the non-crystallization of the graphite occurred, but, beyond an annealing time, recrystallizations of the graphite arise. Higher annealing temperature supported the recrystallization. The milled and annealed expanded graphite still preserved the crystalline structure as raw material and hold high thermal stability.

  6. LBNF 1.2 MW TARGET: CONCEPTUAL DESIGN & FABRICATION

    Energy Technology Data Exchange (ETDEWEB)

    Crowley, Cory F. [Fermilab; Ammigan, K. [Fermilab; Anderson, K. [Fermilab; Hartsell, B. [Fermilab; Hurh, P. [Fermilab; Hylen, J. [Fermilab; Zwaska, R. [Fermilab

    2015-06-29

    Fermilab’s Long-Baseline Neutrino Facility (LBNF) will utilize a modified design based on the NuMI low energy target that is reconfigured to accommodate beam operation at 1.2 MW. Achieving this power with a graphite target material and ancillary systems originally rated for 400 kW requires several design changes and R&D efforts related to material bonding and electrical isolation. Target cooling, structural design, and fabrication techniques must address higher stresses and heat loads that will be present during 1.2 MW operation, as the assembly will be subject to cyclic loads and thermal expansion. Mitigations must be balanced against compromises in neutrino yield. Beam monitoring and subsystem instrumentation will be updated and added to ensure confidence in target positioning and monitoring. Remote connection to the target hall support structure must provide for the eventual upgrade to a 2.4 MW target design, without producing excessive radioactive waste or unreasonable exposure to technicians during reconfiguration. Current designs and assembly layouts will be presented, in addition to current findings on processes and possibilities for prototype and final assembly fabrication.

  7. LBNF 1.2 MW Target: Conceptual Design & Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Crowley, C. [Fermilab; Ammigan, K. [Fermilab; Anderson, K. [Fermilab; Hartsell, B. [Fermilab; Hurh, P. [Fermilab; Hylen, J. [Fermilab; Zwaska, R. [Fermilab

    2015-06-01

    Fermilab’s Long-Baseline Neutrino Facility (LBNF) will utilize a modified design based on the NuMI low energy target that is reconfigured to accommodate beam operation at 1.2 MW. Achieving this power with a graphite target material and ancillary systems originally rated for 400 kW requires several design changes and R&D efforts related to material bonding and electrical isolation. Target cooling, structural design, and fabrication techniques must address higher stresses and heat loads that will be present during 1.2 MW operation, as the assembly will be subject to cyclic loads and thermal expansion. Mitigations must be balanced against compromises in neutrino yield. Beam monitoring and subsystem instrumentation will be updated and added to ensure confidence in target positioning and monitoring. Remote connection to the target hall support structure must provide for the eventual upgrade to a 2.4 MW target design, without producing excessive radioactive waste or unreasonable exposure to technicians during reconfiguration. Current designs and assembly layouts will be presented, in addition to current findings on processes and possibilities for prototype and final assembly fabrication.

  8. On target : extreme engineering's tools used for fundamental and high-end drilling operations

    Energy Technology Data Exchange (ETDEWEB)

    Cope, G.

    2006-09-15

    Positioned at the bottom of the drillstring, measurement-while-drilling systems can discern the location of the drillbit and transmit information back to the surface in the form of pulses in the mud system. The technology is costly, time-consuming and provides a limited amount of information. This article presented details of new devices developed by Extreme Engineering Ltd. to communicate drillbit location information from the bottom of the well to the surface in real time. The XPulse is a mud pulse system with continuous directional measuring capacity that allows operators on location to perform fundamental directional drilling operations without using a technician to interpret data. Installed behind the drillbit, the device records azimuth and inclination when the bit is sitting idle. Inclination is measured using a set of 3 axis accelerometers, quartz crystal devices that are sensitive to the Earth's gravitational force. Azimuth is measured with a flux gate magnetometer that senses its orientation to the Earth's magnetic field. The 2 data are computed to give an orientation of the drillbit in three-dimensional space. Extreme Engineering Ltd. has also developed XAct, a new set of tools using acoustic telemetry to transmit data to the surface using piezo ceramic materials. The acoustic energy is transmitted up the steel drillpipe at the rate of 20 bits per second to the surface, where an electro acoustic receiver detects it. The rate is sufficient to transmit large amounts of geological and drilling information to the surface in real time. Designed for high-end applications, the new suite of tools has cost the company almost $30 million to develop. It was concluded that industry response to the new technologies has been positive. 4 figs.

  9. 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

  10. Proposal of a post-prostatectomy clinical target volume based on pre-operative MRI: volumetric and dosimetric comparison to the RTOG guidelines

    International Nuclear Information System (INIS)

    Croke, Jennifer; Maclean, Jillian; Nyiri, Balazs; Li, Yan; Malone, Kyle; Avruch, Leonard; Kayser, Cathleen; Malone, Shawn

    2014-01-01

    Recurrence rates following radiotherapy for prostate cancer in the post-operative adjuvant or salvage setting remain substantial. Previous work from our institution demonstrated that published prostate bed CTV guidelines frequently do not cover the pre-operative MRI defined prostate. Inadequate target delineation may contribute to the high recurrence rates, but increasing target volumes may increase dose to organs at risk. We propose guidelines for delineating post-prostatectomy target volumes based upon an individual’s co-registered pre-operative MRI. MRI-based CTVs and PTVs were compared to those created using the RTOG guidelines in 30 patients. Contours were analysed in terms of absolute volume, intersection volume (Jaccard Index) and the ability to meet the RADICALS and QUANTEC rectal and bladder constraints (tomotherapy IMRT plans with PTV coverage of V98% ≥98%). CTV MRI was a mean of 18.6% larger than CTV RTOG: CTV MRI mean 138 cc (range 72.3 - 222.2 cc), CTV RTOG mean 116.3 cc (range 62.1 - 176.6 cc), (p < 0.0001). The difference in mean PTV was only 4.6%: PTV MRI mean 386.9 cc (range 254.4 – 551.2), PTV RTOG mean 370 cc (range 232.3 - 501.6) (p = 0.05). The mean Jaccard Index representing intersection volume between CTVs was 0.72 and 0.84 for PTVs. Both criteria had a similar ability to meet rectal and bladder constraints. Rectal DVH: 77% of CTV RTOG cases passed all RADICALS criteria and 37% all QUANTEC criteria; versus 73% and 40% for CTV MRI (p = 1.0 for both). Bladder DVH; 47% of CTV RTOG cases passed all RADICALS criteria and 67% all QUANTEC criteria, versus 57% and 60% for CTV MRI (p = 0.61for RADICALS, p = 0.79 for QUANTEC). CTV MRI spares more of the lower anterior bladder wall than CTV RTOG but increases coverage of the superior lateral bladder walls. CTV contours based upon the patient’s co-registered pre-operative MRI in the post-prostatectomy setting may improve coverage of the individual’s prostate bed without substantially increasing

  11. In-situ electric resistance measurements and annealing effects of graphite exposed to swift heavy ions

    International Nuclear Information System (INIS)

    Fernandes, Sandrina; Pellemoine, Frederique; Tomut, Marilena; Avilov, Mikhail; Bender, Markus; Boulesteix, Marine; Krause, Markus; Mittig, Wolfgang; Schein, Mike; Severin, Daniel; Trautmann, Christina

    2013-01-01

    To study the suitability of using graphite as material for high-power targets for rare isotope production at the future Facility for Rare Isotope Beams (FRIB) in the USA and at the Facility for Antiproton and Ion Research (FAIR) in Germany, thin foils of polycrystalline graphite were exposed to 8.6-MeV/u Au ions reaching a maximum fluence of 1 × 10 15 ions/cm 2 . Foil irradiation temperatures of up to 1800 °C were obtained by ohmic heating. In-situ monitoring of the electrical resistance of the graphite foils during and after irradiation provided information on beam-induced radiation damage. The rate of electrical resistance increase as a function of fluence was found to decrease with increasing irradiation temperature, indicating a more efficient annealing of the irradiation-produced defects. This is corroborated by the observation that graphite foils irradiated at temperatures below about 800 °C showed cracks and pronounced deformations, which did not appear on the samples irradiated at higher temperatures

  12. In-situ electric resistance measurements and annealing effects of graphite exposed to swift heavy ions

    Energy Technology Data Exchange (ETDEWEB)

    Fernandes, Sandrina [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI (United States); Pellemoine, Frederique, E-mail: pellemoi@frib.msu.edu [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI (United States); Tomut, Marilena [GSI Helmholtzzentrum für Schwerionenforschung, Darmstadt (Germany); National Institute for Materials Physics (NIMP), Bucharest (Romania); Avilov, Mikhail [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI (United States); Bender, Markus [GSI Helmholtzzentrum für Schwerionenforschung, Darmstadt (Germany); Boulesteix, Marine [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI (United States); Krause, Markus [GSI Helmholtzzentrum für Schwerionenforschung, Darmstadt (Germany); Technische Universität, Darmstadt (Germany); Mittig, Wolfgang [National Superconducting Cyclotron Lab (NSCL), Michigan State University, East Lansing, MI (United States); Schein, Mike [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI (United States); Severin, Daniel [GSI Helmholtzzentrum für Schwerionenforschung, Darmstadt (Germany); Trautmann, Christina [GSI Helmholtzzentrum für Schwerionenforschung, Darmstadt (Germany); Technische Universität, Darmstadt (Germany)

    2013-11-01

    To study the suitability of using graphite as material for high-power targets for rare isotope production at the future Facility for Rare Isotope Beams (FRIB) in the USA and at the Facility for Antiproton and Ion Research (FAIR) in Germany, thin foils of polycrystalline graphite were exposed to 8.6-MeV/u Au ions reaching a maximum fluence of 1 × 10{sup 15} ions/cm{sup 2}. Foil irradiation temperatures of up to 1800 °C were obtained by ohmic heating. In-situ monitoring of the electrical resistance of the graphite foils during and after irradiation provided information on beam-induced radiation damage. The rate of electrical resistance increase as a function of fluence was found to decrease with increasing irradiation temperature, indicating a more efficient annealing of the irradiation-produced defects. This is corroborated by the observation that graphite foils irradiated at temperatures below about 800 °C showed cracks and pronounced deformations, which did not appear on the samples irradiated at higher temperatures.

  13. Nucleation and growth characteristics of graphite spheroids in bainite during graphitization annealing of a medium carbon steel

    International Nuclear Information System (INIS)

    Gao, J.X.; Wei, B.Q.; Li, D.D.; He, K.

    2016-01-01

    The evolution of microstructure in bainite during graphitization annealing at 680 °C of Jominy-quenched bars of an Al-Si bearing medium carbon (0.4C wt%) steel has been studied and compared with that in martensite by using light, scanning and transmission electron microscopy. The results show that the graphitization process in bainite is different from that in martensite in many aspects such as the initial carbon state, the behavior of cementite, the nucleation-growth feature and kinetics of formation of graphite spheroids during graphitization annealing, and the shape, size and distribution of these graphite spheroids. The fact that the graphitization in bainite can produce more homogeneous graphite spheroids with more spherical shape and finer size in a shorter annealing time without the help of preexisting coring particles implies that bainite should be a better starting structure than martensite for making graphitic steel. - Highlights: • This article presents a microstructural characterization of formation of graphite spheroids in bainite. • Nucleation and growth characteristics of graphite spheroids formed in bainite and martensite are compared. • Bainite should be a better starting structure for making graphitic steel as results show.

  14. Erosion of pyrolytic graphite and Ti-doped graphite due to high flux irradiation

    International Nuclear Information System (INIS)

    Ohtsuka, Yusuke; Ohashi, Junpei; Ueda, Yoshio; Isobe, Michiro; Nishikawa, Masahiro

    1997-01-01

    The erosion of pyrolytic graphite and titanium doped graphite RG-Ti above 1,780 K was investigated by 5 keV Ar beam irradiation with the flux from 4x10 19 to 1x10 21 m -2 ·s -1 . The total erosion yields were significantly reduced with the flux. This reduction would be attributed to the reduction of RES (radiation enhanced sublimation) yield, which was observed in the case of isotropic graphite with the flux dependence of RES yield of φ -0.26 (φ: flux) obtained in our previous work. The yield of pyrolytic graphite was roughly 30% higher than that of isotropic graphite below the flux of 10 20 m -2 ·s -1 whereas each yield approached to very close value at the highest flux of 1x10 21 m -2 ·s -1 . This result indicated that the effect of graphite structure on the RES yield, which was apparent in the low flux region, would disappear in the high flux region probably due to the disordering of crystal structure. In the case of irradiation to RG-Ti at 1,780 K, the surface undulations evolved with a mean height of about 3 μm at 1.2x10 20 m -2 ·s -1 , while at higher flux of 8.0x10 20 m -2 ·s -1 they were unrecognizable. These phenomena can be explained by the reduction of RES of graphite parts excluding TiC grains. (author)

  15. Comprehensive Study for an Optimized Redesign of the CERN's Antiproton Decelerator Target

    CERN Document Server

    AUTHOR|(CDS)2089345; Perillo-Marcone, Antonio; Muñoz-Cobo, Jose-Luis

    2018-04-16

    The Antiproton Decelerator Target (AD-Target) is a unique device responsible for the production of antiprotons at the European Organization for Nuclear Research (CERN). During operation, intense 26 GeV energy proton beams are impacted into its core, made of a 3 mm diameter rod of a high density material such as iridium, creating secondary particles -including antiprotons- from the nuclear reactions induced in its interior. This thesis delves into the characteristics of antiproton production and in particular in the mechanical response of the target core material, which is exposed to a rise of temperature of approximate 2000 degrees Celsius in less than 0.5 microseconds each time is impacted by the primary proton beam. A coupled numerical-experimental approach has been applied for this purpose. Specific computational tools, called hydrocodes, have been used for simulating the extreme dynamic response taking place in the target core and its containing graphite matrix, indicating their potential damage and frag...

  16. A pupal transcriptomic screen identifies Ral as a target of store-operated calcium entry in Drosophila neurons.

    Science.gov (United States)

    Richhariya, Shlesha; Jayakumar, Siddharth; Abruzzi, Katharine; Rosbash, Michael; Hasan, Gaiti

    2017-02-14

    Transcriptional regulation by Store-operated Calcium Entry (SOCE) is well studied in non-excitable cells. However, the role of SOCE has been poorly documented in neuronal cells with more complicated calcium dynamics. Previous reports demonstrated a requirement for SOCE in neurons that regulate Drosophila flight bouts. We refine this requirement temporally to the early pupal stage and use RNA-sequencing to identify SOCE mediated gene expression changes in the developing Drosophila pupal nervous system. Down regulation of dStim, the endoplasmic reticular calcium sensor and a principal component of SOCE in the nervous system, altered the expression of 131 genes including Ral, a small GTPase. Disruption of Ral function in neurons impaired flight, whereas ectopic expression of Ral in SOCE-compromised neurons restored flight. Through live imaging of calcium transients from cultured pupal neurons, we confirmed that Ral does not participate in SOCE, but acts downstream of it. These results identify neuronal SOCE as a mechanism that regulates expression of specific genes during development of the pupal nervous system and emphasizes the relevance of SOCE-regulated gene expression to flight circuit maturation.

  17. 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

  18. 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

  19. Electrochemical behaviour of rhenium-graphite electrode

    International Nuclear Information System (INIS)

    Varypaev, V.N.; Krasikov, V.L.

    1980-01-01

    Electrochemical behaviour of combination electrode from graphite with electrodeposited thin coating of electrolytic rhenium is studied. Solution of 0.5 m NaCl+0.04 m AlCl 3 served as an electrolite. Polarization galvanostatic curves of hydrogen evolution upon electrodes with conditional rhenium thickness of 3.5 and 0.35 μm, 35 and 3.5 nm are obtained. Possibility of preparation of rhenium-graphite cathode with extremely low rhenium consume, electro-chemical properties of which are simu-lar to purely rhenium cathode is shown. Such electrode is characterized with stable in time low cathode potential of hydrogen evolution in chloride electrolyte and during cathode polarization it is not affected by corrosion

  20. 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)

  1. Magnetic response of certain curved graphitic geometries

    International Nuclear Information System (INIS)

    Wang, L.; Davids, P.S.; Saxena, A.; Bishop, A.R.

    1992-01-01

    The quasi-particle energy spectra associated with some members of buckyfamily (curved graphitic geometries), in particular C 50 , C 60 , C 70 and related fullerenes as well as coaxial helical microtubules of graphite, are obtained analytically within the mean-field approximation. These energy spectra are then used to calculate various response functions. Specifically, we calculate the specific heat, magnetization and magnetic susceptibility in the presence of an external magnetic field at low temperatures. For a single microtubule an extra peak superimposed on the first de Haas van Alphen (dHvA) oscillation in magnetic susceptibility is found in the 50--170 Tesla range depending on the radius which is possibly accessible in special (explosive flux compression) experiments. Finally, we point to important potential applications of these novel mesoscopic structures in nanotechnology

  2. Imaging of tritium implanted into graphite

    International Nuclear Information System (INIS)

    Malinowski, M.E.; Causey, R.A.

    1988-01-01

    The extensive use of graphite in plasma-facing surfaces of tokamaks such as the Tokamak Fusion Test Reactor, which has planned tritium discharges, makes two-dimensional tritium detection techniques important in helping to determine torus tritium inventories. We have performed experiments in which highly oriented pyrolytic graphite (HOPG) samples were first tritium implanted with fluences of ∼10 16 T/cm 2 at energies approx. 0 C resulted in no discernible motion of tritium along the basal plane, but did show that significant desorption of the implanted tritium occurred. The current results indicate that tritium in quantities of 10 12 T/cm 2 in tritiated components could be readily detected by imaging at lower magnifications

  3. Preparation of nanoporous carbons from graphite nanofibres

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byung-Joo [Department of Green Chemistry and Environmental Biotechnology, University of Science and Technology, PO Box 107, Yuseong, Daejeon 305-600 (Korea, Republic of); Park, Soo-Jin [Department of Chemistry, Inha University, 253, Nam-gu, Incheon 402-751 (Korea, Republic of)

    2006-09-14

    In this study we manufactured highly porous graphite nanofibres (GNFs) by physical activation in order to develop promising energy storage materials. The activation was performed at activation temperatures in the range of 800-1050 deg. C. The pore structures of the porous GNFs were analysed using N{sub 2}/77 K adsorption isotherms. After the activation, the porous GNFs showed a decrease in diameter and scratches on their surfaces, resulting from surface oxidation and the opening of the graphitic layers, respectively. It was found that the specific surface area of the porous GNFs prepared at 1050 deg. C was more than 2000 m{sup 2} g{sup -1} without loss of their fibre shape or serious increase in electrical resistivity. This result indicates that porous GNFs prepared under optimal conditions can have a much higher specific surface area and are promising materials for energy storage technologies.

  4. Spectroscopical determination of impurities in nuclear graphite

    International Nuclear Information System (INIS)

    Lordello, A.R.; Tognini, R.P.

    1975-01-01

    A spectrochemical method for the direct determination of B, Cd, Si, Hg, Fe, Mg, Mn, Cr, Ni, Al, Mo, Ti, Sr, Na, Zn, and As in nuclear grade graphite is described. A 9:1 ratio of graphite to copper difluoride is used in the preparation of samples and standards. The excitation is carried out in a d-c at 10 amperes. The copper fluoride used as spectrographic buffer serves to increase the volatilization rate of the impurities and to diminish the differences in the nature of the analytical and calibration samples. The relative standard deviations for the determination of the 16 trace elements, except Sr, Fe, Cd, Al and Si, are in the range of 8 - 20% in their appropriate calibration levels. For the latter five elements they are approximately 20-40%

  5. Effect of total pressure on graphite oxidation

    International Nuclear Information System (INIS)

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

    1983-04-01

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

  6. 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

  7. Plastic strain caused by contraction of pores in polycrystalline graphites

    International Nuclear Information System (INIS)

    Ioka, Ikuo; Yoda, Shinichi; Konishi, Takashi.

    1989-01-01

    The effects of porosity on mechanical properties and deformation behavior of four isotropic polycrystalline graphites were studied. The pore size distributions of the graphites were measured using a conventional mercury penetration technique. The average pore radius of ISO-88 graphite was about one-tenth of that of ISEM-1, IG-11 or IG-15 graphites. Young's modulus of the graphites decreased with increasing porosity. The stress-strain curve of each graphite was measured in its lateral and axial directions. Young's modulus of graphite decreased with increasing load. The plastic strain at a given compressive load was calculated from the stress-strain curve and the initial gradient of the unloading curve at the load. The ratio of lateral plastic strain to axial plastic strain for the graphites was less than 0.5, indicating that the volume of the graphites decreased during compressive loading. By assuming that the volume change was caused by contraction of pores, plastic strain associated with contraction of pores was calculated from the axial plastic strain and lateral plastic strain by slips along the basal planes. The plastic strain increased with increasing axial plastic strain and porosity of graphite. (author)

  8. Effect of graphite reflector on activation of fusion breeding blanket

    International Nuclear Information System (INIS)

    Lee, Cheol Woo; Lee, Young-Ouk; Lee, Dong Won; Cho, Seungyon; Ahn, Mu-Young

    2016-01-01

    Highlights: • The graphite reflector concept has been applied in the design of the Korea HCCR TBM for ITER and this concept is also a candidate design option for Korea Demo. • In the graphite reflector, C-14, B-11 and Be-10 are produced after an irradiation. Impurities in both case of beryllium and graphite is dominant in the shutdown dose after an irradiation. • Based on the evaluation, the graphite reflector is a good alternative of the beryllium multiplier in the view of induced activity and shutdown dose. But C-14 produced in the graphite reflector should be considered carefully in the view of radwaste management. - Abstract: Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. In this paper, activity analysis was performed and the effect of graphite reflector in the view of activation was compared to the beryllium multiplier. As a result, it is expected that using the graphite reflector instead of the beryllium multiplier decreases total activity very effectively. But the graphite reflector produces C-14 about 17.2 times than the beryllium multiplier. Therefore, C-14 produced in the graphite reflector is expected as a significant nuclide in the view of radwaste management.

  9. Theoretical analysis of the graphitization of a nanodiamond

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, S Joon; Park, Jae-Gwan [Nano Science and Technology Division, Korea Institute of Science and Technology (KIST), PO Box 131, Cheongryang, Seoul, 130-650 (Korea, Republic of)

    2007-09-26

    We report on a theoretical analysis of the graphitization of a nanosize diamond (nanodiamond) in the metastable state. A nanodiamond annealed at a relatively lower temperature suffers morphological transition into a nanodiamond-graphite core-shell structure. Thermodynamic stability analysis of the nanodiamond showed that the phase diagram (relationship between the annealing temperature and radius) of the nanodiamond-graphite has three regimes: smaller nanodiamond, nanodiamond-graphite, and larger nanodiamond. These regimes of nanodiamond-graphite are due to an additional phase boundary from finding the maximum size of the nanodiamond which can be graphitized. In the theoretical analysis, the most probable and the maximum volume fractions of graphite in the nanodiamond were 0.76 and 0.84 respectively, which were independent of the annealing temperature and the initial radius of the nanodiamond. Therefore, the nanodiamond is not completely transformed into graphite by simple annealing at relatively lower process temperature and pressure. The highest graphitization probability decreased with increasing annealing temperature. Raman spectra for the F{sub 2g} vibration mode of nanodiamond were also calculated, and we found that the variation in properties of the spectral line was strongly dependent on the graphitization temperature and the initial size of the nanodiamond.

  10. Theoretical analysis of the graphitization of a nanodiamond

    International Nuclear Information System (INIS)

    Kwon, S Joon; Park, Jae-Gwan

    2007-01-01

    We report on a theoretical analysis of the graphitization of a nanosize diamond (nanodiamond) in the metastable state. A nanodiamond annealed at a relatively lower temperature suffers morphological transition into a nanodiamond-graphite core-shell structure. Thermodynamic stability analysis of the nanodiamond showed that the phase diagram (relationship between the annealing temperature and radius) of the nanodiamond-graphite has three regimes: smaller nanodiamond, nanodiamond-graphite, and larger nanodiamond. These regimes of nanodiamond-graphite are due to an additional phase boundary from finding the maximum size of the nanodiamond which can be graphitized. In the theoretical analysis, the most probable and the maximum volume fractions of graphite in the nanodiamond were 0.76 and 0.84 respectively, which were independent of the annealing temperature and the initial radius of the nanodiamond. Therefore, the nanodiamond is not completely transformed into graphite by simple annealing at relatively lower process temperature and pressure. The highest graphitization probability decreased with increasing annealing temperature. Raman spectra for the F 2g vibration mode of nanodiamond were also calculated, and we found that the variation in properties of the spectral line was strongly dependent on the graphitization temperature and the initial size of the nanodiamond

  11. Graphite-graphite oxide composite electrode for vanadium redox flow battery

    International Nuclear Information System (INIS)

    Li Wenyue; Liu Jianguo; Yan Chuanwei

    2011-01-01

    Highlights: → A new composite electrode is designed for vanadium redox flow battery (VRB). → The graphite oxide (GO) is used as electrode reactions catalyst. → The excellent electrode activity is attributed to the oxygen-containing groups attached on the GO surface. → A catalytic mechanism of the GO towards the redox reactions is presumed. - Abstract: A graphite/graphite oxide (GO) composite electrode for vanadium redox battery (VRB) was prepared successfully in this paper. The materials were characterized with X-ray diffraction, X-ray photoelectron spectroscopy and scanning electron microscopy. The specific surface area was measured by the Brunauer-Emmett-Teller method. The redox reactions of [VO 2 ] + /[VO] 2+ and V 3+ /V 2+ were studied with cyclic voltammetry and electrochemical impedance spectroscopy. The results indicated that the electrochemical performances of the electrode were improved greatly when 3 wt% GO was added into graphite electrode. The redox peak currents of [VO 2 ] + /[VO] 2+ and V 3+ /V 2+ couples on the composite electrode were increased nearly twice as large as that on the graphite electrode, and the charge transfer resistances of the redox pairs on the composite electrode are also reduced. The enhanced electrochemical activity could be ascribed to the presence of plentiful oxygen functional groups on the basal planes and sheet edges of the GO and large specific surface areas introduced by the GO.

  12. Determining the future for irradiated graphite disposal

    International Nuclear Information System (INIS)

    Neighbour, G.B.; Wickham, A.J.; Hacker, P.J.

    2000-01-01

    In recent years, proposals have been made for the long-term treatment of radioactive graphite waste which have ranged from sea dumping through incineration to land-based disposal, sometimes preceded by a variable period of 'safe storage' within the original reactor containment. Nuclear regulators are challenging the proposed length of 'safe storage' on the basis that essential knowledge may be lost. More recently, political constraints have further complicated the issue by eliminating disposal at sea and imposing a 'near-zero release' philosophy, while public opinion is opposed to land-based disposal and has induced a continual drive towards minimizing radioactivity release to the environment from disposal. This paper proposes that, despite various international agreements, it is time to review technically all options for disposal of irradiated graphite waste as a framework for the eventual decision-making process. It is recognized that the socio-economic and political pressures are high and therefore, given that all currently identified options satisfy the present safety limits, the need to minimize the objective risk is shown to be a minor need in comparison to the public's want of demonstrable control, responsiveness and ability to reverse/change the disposal options in the future. Further, it is shown that the eventual decision-making process for a post-dismantling option for graphite waste must optimize the beneficial attributes of subjective risk experienced by the general public. In addition, in advocating and preferred option to the general public, it is recommended that the industry should communicate at a level commensurate with the public understanding and initiate a process of facilitation which enables the public to arrive at their own solution and constituting a social exchange. Otherwise it is concluded that if the indecision over disposal options is allowed to continue then, by default, graphite will remain in long-term supervised storage. (author)

  13. Graphite-to-Graphene: Total Conversion.

    Science.gov (United States)

    Buzaglo, Matat; Bar, Ilan Pri; Varenik, Maxim; Shunak, Liran; Pevzner, Svetlana; Regev, Oren

    2017-02-01

    The rush to develop graphene applications mandates mass production of graphene sheets. However, the currently available complex and expensive production technologies are limiting the graphene commercialization. The addition of a protective diluent to graphite during ball-milling is demonstrated to result in a game-changer yield (>90%) of defect-free graphene, whose size is controlled by the milling energy and the diluent type. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  14. Electron oxidation of graphite by fluorospecies

    International Nuclear Information System (INIS)

    Rosenthal, G.L.

    1984-09-01

    The fluoride-ion affinity (A/sub F - /) of phosphorus pentafluoride was determined to be 100 kcal/mole from the heats of reaction of the Lewis bases SF 4 and ClO 2 F with PF 5 near room temperature. The fluoride-ion affinity of boron trifluoride was determined to be 92 kcal/mole from the heat of reaction of ClO 2 F with BF 3 . The crystal structure of ClO 2 BF 4 was determined and a precise lattice energy was calculated from this structure and used to determined A/sub F - /. Both PF 5 and BF 3 were found to react with graphite in the presence of fluorine gas to yield a variety of non-stoichiometric compounds. The fluoride-ion affinity of silicon tetrafluoride is not known, but it does not react with graphite and F 2 except at high pressures. These and previous results suggested a threshold in oxidizing power of intercalating species below which the oxidative intercalation reaction would not occur. The reduction of C/sub x/PF 6 by PF 3 proved that the reaction is thermodynamically controlled to some extent. The displacement of PF 5 in C/sub x/PF 6 by BF 3 (with a smaller A/sub F - /) suggested that two BF 3 molecules may have a larger fluoride-ion affinity than one PF 5 and that B 2 F 7 - may be a stable anion in graphite. Conductivity studies of PF/sub x/ and BF/sub y/ salts showed that a large drop in conductivity when the reaction reaches first stage is due in the most part to direct fluorination of carbon in graphite

  15. Temperature dependence of phonons in pyrolitic graphite

    International Nuclear Information System (INIS)

    Brockhouse, B.N.; Shirane, G.

    1977-01-01

    Dispersion curves for longitudinal and transverse phonons propagating along and near the c-axis in pyrolitic graphite at temperatures between 4 0 K and 1500 0 C have been measured by neutron spectroscopy. The observed frequencies decrease markedly with increasing temperature (except for the transverse optical ''rippling'' modes in the hexagonal planes). The neutron groups show interesting asymmetrical broadening ascribed to interference between one phonon and many phonon processes

  16. High thermoelectric performance of graphite nanofibers

    OpenAIRE

    Tran, Van-Truong; Saint-Martin, Jérôme; Dollfus, Philippe; Volz, Sebastian

    2017-01-01

    Graphite nanofibers (GNFs) have been demonstrated to be a promising material for hydrogen storage and heat management in electronic devices. Here, by means of first-principles and transport simulations, we show that GNFs can also be an excellent material for thermoelectric applications thanks to the interlayer weak van der Waals interaction that induces low thermal conductance and a step-like shape in the electronic transmission with mini-gaps, which are necessary ingredients to achieve high ...

  17. Increase of the density of commercial graphite

    International Nuclear Information System (INIS)

    Tobias, H.; Meyerstein, D.

    1977-12-01

    The increase of the density of commercial graphite of the type ATJ by polymerization of an impregnated monomer, followed by pyrolysis, is described. The monomer which was either styrene or acrylonitrile, was irradiated by a 60 Co source and pyrolized in a standard vacuum system. The irradiation dose for the polymerization of the monomer was determined. Suggestions for the establishment of the optimum conditions are offered

  18. 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.

  19. Actinides in irradiated graphite of RBMK-1500 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Plukienė, R., E-mail: rita@ar.fi.lt; Plukis, A.; Barkauskas, V.; Gudelis, A.; Gvozdaitė, R.; Duškesas, G.; Remeikis, V.

    2014-10-01

    Highlights: • Activation of actinides in the graphite of the RBMK-1500 reactor was analyzed. • Numerical modeling using SCALE 6.1 and MCNPX was used for actinide calculation. • Measurements of the irradiated graphite sample were used for model validation. • Results are important for further decommissioning process of the RBMK type reactors. - Abstract: The activation of graphite in the nuclear power plants is the problem of high importance related with later graphite reprocessing or disposal. The activation of actinide impurities in graphite due to their toxicity determines a particular long term risk to waste management. In this work the activation of actinides in the graphite constructions of the RBMK-1500 reactor is determined by nuclear spectrometry measurements of the irradiated graphite sample from the Ignalina NPP Unit I and by means of numerical modeling using two independent codes SCALE 6.1 (using TRITON-VI sequence) and MCNPX (v2.7 with CINDER). Both models take into account the 3D RBMK-1500 reactor core fragment with explicit graphite construction including a stack and a sleeve but with a different simplification level concerning surrounding graphite and construction of control roads. The verification of the model has been performed by comparing calculated and measured isotope ratios of actinides. Also good prediction capabilities of the actinide activation in the irradiated graphite have been found for both calculation approaches. The initial U impurity concentration in the graphite model has been adjusted taking into account the experimental results. The specific activities of actinides in the irradiated RBMK-1500 graphite constructions have been obtained and differences between numerical simulation results, different structural parts (sleeve and stack) as well as comparison with previous results (Ancius et al., 2005) have been discussed. The obtained results are important for further decommissioning process of the Ignalina NPP and other RBMK

  20. 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.