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)

Utilizing the slowing-down-time technique for benchmarking neutron thermalization in graphite

International Nuclear Information System (INIS)

Zhou, T.; Hawari, A. I.; Wehring, B. W.

2007-01-01

Graphite is the moderator/reflector in the Very High Temperature Reactor (VHTR) concept of Generation IV reactors. As a thermal reactor, the prediction of the thermal neutron spectrum in the VHTR is directly dependent on the accuracy of the thermal neutron scattering libraries of graphite. In recent years, work has been on-going to benchmark and validate neutron thermalization in 'reactor grade' graphite. Monte Carlo simulations using the MCNP5 code were used to design a pulsed neutron slowing-down-time experiment and to investigate neutron slowing down and thermalization in graphite at temperatures relevant to VHTR operation. The unique aspect of this experiment is its ability to observe the behavior of neutrons throughout an energy range extending from the source energy to energies below 0.1 eV. In its current form, the experiment is designed and implemented at the Oak Ridge Electron Linear Accelerator (ORELA). Consequently, ORELA neutron pulses are injected into a 70 cm x 70 cm x 70 cm graphite pile. A furnace system that surrounds the pile and is capable of heating the graphite to a centerline temperature of 1200 K has been designed and built. A system based on U-235 fission chambers and Li-6 scintillation detectors surrounds the pile. This system is coupled to multichannel scaling instrumentation and is designed for the detection of leakage neutrons as a function of the slowing-down-time (i.e., time after the pulse). To ensure the accuracy of the experiment, careful assessment was performed of the impact of background noise (due to room return neutrons) and pulse-to-pulse overlap on the measurement. Therefore, the entire setup is surrounded by borated polyethylene shields and the experiment is performed using a source pulse frequency of nearly 130 Hz. As the basis for the benchmark, the calculated time dependent reaction rates in the detectors (using the MCNP code and its associated ENDF-B/VI thermal neutron scattering libraries) are compared to measured

Thermal shock test of TiC and graphite

International Nuclear Information System (INIS)

Shirakawa, H.; Okamura, J.; Son, P.; Miyake, M.

1989-01-01

Thermal shock tests were performed by pulse electron beam heating on chemically vapor deposited coatings of TiC on Poco graphite, bulk TiC, and several kinds of isotropic graphite. The specimens were heated at various power densities (10-45 MW/m 2 ) for various pulse durations (1-2 s) to examine the dependence of thermal failures on heating conditions. The TiC coating on graphite suffered cracking, surface melting and evaporation by the thermal pulse. The surface melting limit, defined as F τ 1/2 , where F is the minimum power density that causes surface melting for a specified pulse duration τ, was approximately 48 MWs 1/2 /m 2 for the TiC coating. The combined-Carbon/Titanium ratio of the coating after electron beam heating decreased with increasing power density and pulse duration. The bulk TiC specimens were so brittle that they fractured at heat load conditions where the coating showed no damage. The graphite specimens showed sublimation as a principal damage mechanism by the thermal pulse, and the sublimation weight loss decreased with increasing the thermal conductivity of the specimen. It was confirmed that the TiC coating on graphite had favorable resistance to thermal shock as compared to the bulk TiC and that graphite with high thermal conductivity is promising material as a high heat flux component. (orig.)

Calculation of reactivity of control rods in graphite moderated reactors

International Nuclear Information System (INIS)

Nakata, H.

1978-01-01

A study about the method of calculation for the reactivity of control rods in graphite-moderated critical assemblies, is presented. The result of theoretical calculation, developed by super celles and Nordheim-Scalettar methods are compared with experimental results for the critical Assembly of General Atomic. The two methods are then applicable to reactivity calculation of the control rods of graphite moderated critical assemblies [pt

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-02-20

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

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.

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

Study on thermal conductivity of HTR spherical fuel element matrix graphite

International Nuclear Information System (INIS)

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

2014-01-01

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

Effective Thermal Conductivity of Graphite Materials with Cracks

Science.gov (United States)

Pestchaanyi, S. E.; Landman, I. S.

The dependence of effective thermal diffusivity on temperature caused by volumetric cracks is modelled for macroscopic graphite samples using the three-dimensional thermomechanics code Pegasus-3D. At high off-normal heat loads typical of the divertor armour, thermostress due to the anisotropy of graphite grains is much larger than that due to the temperature gradient. Numerical simulation demonstrated that the volumetric crack density both in fine grain graphites and in the CFC matrix depends mainly on the local sample temperature, not on the temperature gradient. This allows to define an effective thermal diffusivity for graphite with cracks. The results obtained are used to explain intense cracking and particle release from carbon based materials under electron beam heat load. Decrease of graphite thermal diffusivity with increase of the crack density explains particle release mechanism in the experiments with CFC where a clear energy threshold for the onset of particle release has been observed in J. Linke et al. Fusion Eng. Design, in press, Bazyler et al., these proceedings. Surface temperature measurement is necessary to calibrate the Pegasus-3D code for simulation of ITER divertor armour brittle destruction.

Thermal migration of deuterium implanted in graphite: Influence of free surface proximity and structure

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, F-69622 Villeurbanne cedex (France); Moncoffre, N., E-mail: n.moncoffre@ipnl.in2p3.fr [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon, F-69622 Villeurbanne cedex (France); Toulhoat, N. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon, 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, F-69622 Villeurbanne cedex (France); Institut Universitaire Technologique, Université Claude Bernard Lyon 1, Université de Lyon, F-69622 Villeurbanne cedex (France); Ammar, M.R. [CNRS, CEMHTI UPR3079, Université Orléans, CS90055, F-45071 Orléans cedex 2 (France); Rouzaud, J.N.; Deldicque, D. [Laboratoire de Géologie de l’Ecole Normale Supérieure, Paris, UMR CNRS ENS 8538, F-75231 Paris cedex 5 (France)

2016-03-15

This paper is a contribution to the study of the behavior of activation products produced in irradiated nuclear graphite, graphite being the moderator of the first French generation of CO{sub 2} cooled nuclear fission reactors. This paper is focused on the thermal release of Tritium, a major contributor to the initial activity, taking into account the role of the free surfaces (open pores and graphite surface). Two kinds of graphite were compared. On one hand, Highly Oriented Pyrolitic Graphite (HOPG), a model well graphitized graphite, and on the other hand, SLA2, a porous less graphitized nuclear graphite. Deuterium ion implantation at three different energies 70, 200 and 390 keV allows simulating the presence of Tritium at three different depths, corresponding respectively to projected ranges R{sub p} of 0.75, 1.7 and 3.2 μm. The D isotopic tracing is performed thanks to the D({sup 3}He,p){sup 4}He nuclear reaction. The graphite structure is studied by Raman microspectrometry. Thermal annealing is performed in the temperature range 200–1200 °C up to 300 h annealing time. As observed in a previous study, the results show that the D release occurs according to three kinetic regimes: a rapid permeation through open pores, a transient regime corresponding to detrapping and diffusion of D located at low energy sites correlated to the edges of crystallites and finally a saturation regime attributed to detrapping of interstitial D located at high energy sites inside the crystallites. Below 600 °C, D release is negligible whatever the implantation depth and the graphite type. The present paper clearly puts forward that above 600 °C, the D release decreases at deeper implantation depths and strongly depends on the graphite structure. In HOPG where high energy sites are more abundant, the D release is less dependent on the surface proximity compared to SLA2. In SLA2, in which the low energy sites prevail, the D release curves are clearly shifted towards lower

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Effect of Residence Time of Graphitisation on Thermal Conductivity of Molded Graphite

Directory of Open Access Journals (Sweden)

Pedy Artsanti

2010-06-01

Full Text Available The effect of residence time of graphitisation on thermal conductivity of molded graphite has been examined. The examination has been conducted by varying residence time of graphitisation of molded carbon with petroleum coke as raw material and coal tar pitch. Graphitisation has been conducted by heating molded graphite at 2500 °C in argon atmosphere with residention time of 10, 30 and 90 minutes. Graphitisation degree, density, shrinking mass and porosity of molded graphite were examined and so was its thermal conductivity. The result showed that the decrease of porosity and the increase of graphitisation degree due to the increasing of residention time of graphitisation will increase the thermal conductivity of graphite. Molded graphite graphitisized with residence time for 90 minutes residention time gave thermal conductivity of 2.134 Watt/mK and graphitization degree 0.718.

Thermal behaviour of chlorine in nuclear graphite at a microscopic scale

Energy Technology Data Exchange (ETDEWEB)

Vaudey, C.E., E-mail: vaudey@ipnl.in2p3.f [Universite de Lyon, Universite Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucleaire de Lyon (IPNL), 4 rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Toulhoat, N., E-mail: nelly.toulhoat@univ-lyon1.f [Universite de Lyon, Universite Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucleaire de Lyon (IPNL), 4 rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Commissariat a l' Energie Atomique CEA/DEN, Centre de Saclay, F-91191 Gif sur Yvette cedex (France); Moncoffre, N. [Universite de Lyon, Universite Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucleaire de Lyon (IPNL), 4 rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Bererd, N. [Universite de Lyon, Universite Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucleaire de Lyon (IPNL), 4 rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Universite de Lyon, UCBL-IUT A, departement chimie, 43 Bd du 11 novembre 1918, F-69622 Villeurbanne cedex (France); Raimbault, L. [Ecole des Mines de Paris, Centre de Geosciences, 35 rue Saint Honore, F-77305 Fontainebleau cedex (France); Sainsot, P. [Universite de Lyon, Universite Lyon 1, LaMCoS, INSA-Lyon, CNRS UMR5259, F-69621 Villeurbanne cedex (France); Rouzaud, J.N. [Laboratoire de Geologie de l' Ecole Normale Superieure, UMR CNRS-ENS 8538, 24 rue Lhomond, F-75231 Paris cedex 5 (France); Perrat-Mabilon, A. [Universite de Lyon, Universite Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucleaire de Lyon (IPNL), 4 rue Enrico Fermi, F-69622 Villeurbanne cedex (France)

2009-12-15

In this paper, we present the study of thermal behaviour of {sup 36}Cl in nuclear graphite used in the St. Laurent A2 UNGG reactor (graphite moderated and CO{sub 2} cooled reactor). {sup 37}Cl, used to simulate {sup 36}Cl displaced from its original structural site by recoil, has been implanted into nuclear graphite samples (energy = 250 keV, fluence = 5 x 10{sup 13} at cm{sup -2}). The samples have been annealed in the 200-800 deg. C temperature range and analysed by SIMS. Structural modifications have been controlled by Raman microspectroscopy. This study shows that, in the considered temperature range and for a short annealing duration (4 h), chlorine is released almost athermally. At 500 deg. C, around 20% of the initial {sup 37}Cl content is released. At 800 deg. C, the release reaches a plateau and the loss of {sup 37}Cl is around 30%. Raman microspectroscopy shows that {sup 37}Cl implantation induces a structural disorder and that during annealing, the original structure is not completely recovered.

Linear expansion of products out of thermal splitting graphite

International Nuclear Information System (INIS)

Tishina, E.A.; Kurnevich, G.I.

1994-01-01

Linear expansion of thermally split graphite in the form of foil and pressed items of different density was studied. It is ascertained that the extreme character of temperature dependence of linear expansion factor of pressed samples of thermally split graphite is determined by the formation of closed pores containing air in the course of their production. 3 refs., 2 figs

Electrical and thermal properties of graphite/polyaniline composites

Energy Technology Data Exchange (ETDEWEB)

Bourdo, Shawn E., E-mail: sxbourdo@ualr.edu [Center for Integrative Nanotechnology Sciences, University of Arkansas at Little Rock, 2801 South University Avenue, Little Rock, AR 72204 (United States); Warford, Brock A.; Viswanathan, Tito [Department of Chemistry, University of Arkansas at Little Rock, 2801 South University Avenue, Little Rock, AR 72204 (United States)

2012-12-15

A composite of a carbon allotrope (graphite) and an inherently conducting polymer, polyaniline (PANI), has been prepared that exhibits an electrical conductivity greater than either of the two components. An almost 2-fold increase in the bulk conductivity occurs when only a small mass fraction of polyaniline exists in the composite (91% graphite/ 9% polyaniline, by mass). This increase in dc electrical conductivity is curious since in most cases a composite material will exhibit a conductivity somewhere between the two individual components, unless a modification to the electronic nature of the material occurs. In order to elucidate the fundamental electrical properties of the composite we have performed variable temperature conductivity measurements to better understand the nature of conduction in these materials. The results from these studies suggest a change in the mechanism of conduction as the amount of polyaniline is increased in the composite. Along with superior electrical properties, the composites exhibit an increase in thermal stability as compared to the graphite. - Graphical abstract: (Left) Room temperature electrical conductivity of G-PANI composites at different mass ratios. (Right) Electrical conductivity of G-PANI composites at temperatures from 5 K to 300 K. Highlights: Black-Right-Pointing-Pointer Composites of graphite and polyaniline have been synthesized with unique electrical and thermal properties. Black-Right-Pointing-Pointer Certain G-PANI composites are more conductive and more thermally stable than graphite alone. Black-Right-Pointing-Pointer G-PANI composites exhibit a larger conductivity ratio with respect to temperature than graphite alone.

Automotive body panel containing thermally exfoliated graphite oxide

Science.gov (United States)

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

2011-01-01

An automotive body panel containing a polymer composite formed of at least one polymer and a modified graphite oxide material, which is a thermally exfoliated graphite oxide with a surface area of from about 300 m.sup.2/g to 2600 m.sup.2/g.

Study on paraffin/expanded graphite composite phase change thermal energy storage material

International Nuclear Information System (INIS)

Zhang Zhengguo; Fang Xiaoming

2006-01-01

A paraffin/expanded graphite composite phase change thermal energy storage material was prepared by absorbing the paraffin into an expanded graphite that has an excellent absorbability. In such a composite, the paraffin serves as a latent heat storage material and the expanded graphite acts as the supporting material, which prevents leakage of the melted paraffin from its porous structure due to the capillary and surface tension forces. The inherent structure of the expanded graphite did not change in the composite material. The solid-liquid phase change temperature of the composite PCM was the same as that of the paraffin, and the latent heat of the paraffin/expanded graphite composite material was equivalent to the calculated value based on the mass ratio of the paraffin in the composite. The heat transfer rate of the paraffin/expanded graphite composite was obviously higher than that of the paraffin due to the combination with the expanded graphite that had a high thermal conductivity. The prepared paraffin/expanded graphite composite phase change material had a large thermal storage capacity and improved thermal conductivity and did not experience liquid leakage during its solid-liquid phase change

Improvement of thermal shock resistance of isotropic graphite by Ti-doping

International Nuclear Information System (INIS)

Lopez-Galilea, I.; Ordas, N.; Garcia-Rosales, C.; Lindig, S.

2009-01-01

Ti-doped isotropic graphite is a promising candidate material for the strike point area of the ITER divertor due to its reduced chemical erosion by hydrogen bombardment and its high thermal shock resistance, mainly due the catalytic effect of TiC on the graphitization leading to an increase of thermal conductivity and to higher mechanical strength. Several manufacturing parameters such as oxidative stabilization treatment, carbonization cycle, graphitization temperature and dwell time during graphitization have been investigated in order to establish a relationship between these parameters and the final properties.

Improvement of thermal shock resistance of isotropic graphite by Ti-doping

Energy Technology Data Exchange (ETDEWEB)

Lopez-Galilea, I. [Inmaculada Lopez-Galilea, CEIT and Tecnun (University of Navarra), Po de Manuel Lardizabal, 15 E-20018 San Sebastian (Spain)], E-mail: ilopez@ceit.es; Ordas, N.; Garcia-Rosales, C. [Inmaculada Lopez-Galilea, CEIT and Tecnun (University of Navarra), Po de Manuel Lardizabal, 15 E-20018 San Sebastian (Spain); Lindig, S. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, D-85748 Garching (Germany)

2009-04-30

Ti-doped isotropic graphite is a promising candidate material for the strike point area of the ITER divertor due to its reduced chemical erosion by hydrogen bombardment and its high thermal shock resistance, mainly due the catalytic effect of TiC on the graphitization leading to an increase of thermal conductivity and to higher mechanical strength. Several manufacturing parameters such as oxidative stabilization treatment, carbonization cycle, graphitization temperature and dwell time during graphitization have been investigated in order to establish a relationship between these parameters and the final properties.

Critical experiments on enriched uranium graphite moderated cores

International Nuclear Information System (INIS)

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

1978-07-01

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

Enhancing thermal conductivity of fluids with graphite nanoparticles and carbon nanotube

Science.gov (United States)

Zhang, Zhiqiang [Lexington, KY; Lockwood, Frances E [Georgetown, KY

2008-03-25

A fluid media such as oil or water, and a selected effective amount of carbon nanomaterials necessary to enhance the thermal conductivity of the fluid. One of the preferred carbon nanomaterials is a high thermal conductivity graphite, exceeding that of the neat fluid to be dispersed therein in thermal conductivity, and ground, milled, or naturally prepared with mean particle size less than 500 nm, and preferably less than 200 nm, and most preferably less than 100 nm. The graphite is dispersed in the fluid by one or more of various methods, including ultrasonication, milling, and chemical dispersion. Carbon nanotubes with graphitic structure is another preferred source of carbon nanomaterial, although other carbon nanomaterials are acceptable. To confer long term stability, the use of one or more chemical dispersants is preferred. The thermal conductivity enhancement, compared to the fluid without carbon nanomaterial, is proportional to the amount of carbon nanomaterials (carbon nanotubes and/or graphite) added.

Synergistically improved thermal conductivity of polyamide-6 with low melting temperature metal and graphite

Directory of Open Access Journals (Sweden)

Y. C. Jia

2016-08-01

Full Text Available Low melting temperature metal (LMTM-tin (Sn was introduced into polyamide-6 (PA6 and PA6/graphite composites respectively to improve the thermal conductivity of PA6 by melt processing (extruding and injection molding. After introducing Sn, the thermal conductivity of PA6/Sn was nearly constant because of the serious agglomeration of Sn. However, when 20 wt% (5.4 vol% of Sn was added into PA6 containing 50 wt% (33.3 vol% of graphite, the thermal conductivity of the composite was dramatically increased to 5.364 versus 1.852 W·(m·K–1 for the PA6/graphite composite, which suggests that the incorporation of graphite and Sn have a significant synergistic effect on the thermal conductivity improvement of PA6. What is more, the electrical conductivity of the composite increased nearly 8 orders of magnitudes after introducing both graphite and Sn. Characterization of microstructure and energy dispersive spectrum analysis (EDS indicates that the dispersion of Sn in PA6/graphite/Sn was much more uniform than that of PA6/Sn composite. According to Differential Scanning Calorimetry measurement and EDS, the uniform dispersion of Sn in PA6/graphite/Sn and the high thermal conductivity of PA6/graphite/Sn are speculated to be related with the electron transfer between graphite and Sn, which makes Sn distribute evenly around the graphite layers.

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-12-15

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

Graphite moderator annealing of the experimental reactor for irradiation (0.5 MW)

International Nuclear Information System (INIS)

Oliveira Avila, Carlos Alberto de; Pires, Luis Fernando Goncalves

1995-01-01

This work describes an operational procedure for the annealing of the graphite moderator in the 0,5 MW Experimental Reactor for Irradiation. A theoretical methodology has been developed for calculating the temperature field during the annealing process. The equations for mass, momentum, and energy conservation for the coolant as well as for the energy conservation in the moderator are solved numerically. The energy stored in the graphite and released in the annealing is accounted for by the use of a modified source term in the energy conservation equation for the moderator. A good agreement has been found for comparisons of the calculations with annealing data from the BEPO reactor. The major parameters affecting annealing have also been determined. (author). 8 refs, 11 figs

Through-thickness thermal conductivity enhancement of graphite film/epoxy composite via short duration acidizing modification

Science.gov (United States)

Wang, Han; Wang, Shaokai; Lu, Weibang; Li, Min; Gu, Yizhou; Zhang, Yongyi; Zhang, Zuoguang

2018-06-01

Graphite films have excellent in-plane thermal conductivity but extremely low through-thickness thermal conductivity because of their intrinsic inter-layer spaces. To improve the inter-layer heat transfer of graphite films, we developed a simple interfacial modification with a short duration mixed-acid treatment. The effects of the mixture ratio of sulfuric and nitric acids and treatment time on the through-thickness thermal properties of graphite films were studied. The modification increased the through-thickness thermal conductivity by 27% and 42% for the graphite film and its composite, respectively. X-ray photoelectron spectroscopy, X-ray powder diffraction, and scanning electron microscopy results indicated that the acidification process had two competing effects: the positive contribution made by the enhanced interaction between the graphite layers induced by the functional groups and the negative effect from the destruction of the graphite layers. As a result, an optimal acidification method was found to be sulfuric/nitric acid treatment with a mixture ratio of 3:1 for 15 min. The resultant through-thickness thermal conductivity of the graphite film could be improved to 0.674 W/mK, and the corresponding graphite/epoxy composite shows a through-thickness thermal conductivity of 0.587 W/mK. This method can be directly used for graphite films and their composite fabrication to improve through-thickness thermal conductivity.

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

International Nuclear Information System (INIS)

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

1975-01-01

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

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)

Moderator for nuclear reactor

International Nuclear Information System (INIS)

Milgram, M.S.; Dunn, J.T.; Hart, R.S.

1995-01-01

This invention relates to a moderator for a nuclear reactor and more specifically, to a composite moderator. A moderator is designed to slow down, or thermalize, neutrons which are released during nuclear reactions in the reactor fuel. Pure or almost pure materials like light water, heavy water, beryllium or graphite are used singly as moderators at present. All these materials, are used widely. Graphite has a good mechanical strength at high temperatures encountered in the nuclear core and therefore is used as both the moderator and core structural material. It also exhibits a low neutron-capture cross section and high neutron scattering cross section. However, graphite is susceptible to attach by carbon dioxide and/or oxygen where applicable, and releases stress energy under certain circumstances, although under normal operating conditions these reactions can be controlled. (author). 1 tab

KNO3/NaNO3 - Graphite materials for thermal energy storage at high temperature: Part I. - Elaboration methods and thermal properties

International Nuclear Information System (INIS)

Acem, Zoubir; Lopez, Jerome; Palomo Del Barrio, Elena

2010-01-01

Composites graphite/salt for thermal energy storage at high temperature (∼200 deg. C) have been developed and tested. As at low temperature in the past, graphite has been used to enhance the thermal conductivity of the eutectic system KNO 3 /NaNO 3 . A new elaboration method has been proposed as an alternative to graphite foams infiltration. It consists of cold-compression of a physical mixing of expanded natural graphite particles and salt powder. Two different compression routes have been investigated: uni-axial compression and isostatic compression. The first part of the paper has been devoted to the analysis of the thermal properties of these new graphite/salt composites. It is proven that cold-compression is a simple and efficient technique for improving the salt thermal conductivity. For instance, graphite amounts between 15 and 20%wt lead to apparent thermal conductivities close to 20 W/m/K (20 times greater than the thermal conductivity of the salt). Furthermore, some advantages in terms of cost and safety are expected because materials elaboration is carried out at room temperature. The second part of the paper is focused on the analyses of the phase transition properties of these graphite/salt composites materials.

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)

Design of the Graphite Reflectors in Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Lee, Jin Haeng; Cho, Yeong Garp; Kim, Tae Kyu; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-05-15

Graphite is often used as one of reflector materials for research reactors because of its low neutron absorption cross-section, good moderating properties, and relatively low and stable price. In addition, graphite has excellent properties at high temperatures, so it is widely used as a core material in high temperature reactors. However, its material characteristics such as strength, elastic modulus, thermal expansion coefficient, dimensional change, and thermal conductivity sensitively depend on neutron fluence, temperature, and its manufacturing process. In addition, the Wigner energy and the treatment of the graphite waste such as C-14 should also be considered. For the design of the graphite reflectors, it is therefore essential to understand the material characteristics of chosen graphite materials at given conditions. Especially, the dimensional changes and the thermal conductivity are very important factors to design the nuclear components using graphite as a nonstructural material. Hence, in this study, the material characteristics of graphite are investigated via some experiments in literature. Improving design methods for graphite reflectors in research reactors are then suggested to minimize the problems, and the advantages and disadvantages of each method are also discussed

Measurement of the diffusion length of thermal neutrons inside graphite

International Nuclear Information System (INIS)

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

1948-11-01

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

Revisiting the effects of organic solvents on the thermal reduction of graphite oxide

International Nuclear Information System (INIS)

Barroso-Bujans, Fabienne; Fierro, José Luis G.; Alegría, Angel; Colmenero, Juan

2011-01-01

Highlights: ► Retention of organic solvent on graphite oxide interlayer space. ► Decreasing exfoliation temperature. ► Close link between structure and thermal behavior of solvent treated graphite oxide. ► Restacking inhibition of thermally reduced graphite oxide sheets. ► Changes in kinetic mechanisms of thermal reduction. - Abstract: Treatment of graphite oxide (GO) with organic solvents via sorption from either liquid or gas phase, and subsequent desorption, induces profound changes in the layered GO structure: loss of stacking order, retention of trace amounts of solvents and decreasing decomposition temperature. This study presents new evidences of the effect of organic solvents on the thermal reduction of GO by means of thermogravimetric analysis, X-ray diffraction and X-ray photoelectron spectroscopy. The results reveal a relative higher decrease of the oxygen amounts in solvent-treated GO as compared to untreated GO and the restacking inhibition of the thermally reduced GO sheets upon slow heating. The kinetic experiments evidence changes occurring in the reduction mechanisms of the solvent-treated GO, which support the close link between GO structure and thermal properties.

Determination of thermal neutrons diffusion length in graphite

International Nuclear Information System (INIS)

Garcia Fite, J.

1959-01-01

The diffusion length of thermal neutrons in graphite using the less possible quantity of material has been determined. The proceeding used was the measurement in a graphite pile which has a punctual source of rapid neutrons inside surrounded by a reflector medium (paraffin or water). The measurement was done in the following conditions: a) introducing an aluminium plate between both materials. b) Introducing a cadmium plate between both materials. (Author) 91 refs

Packaging material and flexible medical tubing containing thermally exfoliated graphite oxide

Science.gov (United States)

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

2011-01-01

A packaging material or flexible medical tubing 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.

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.

Temperature dependence of the thermal expansion of neutron-irradiated pyrolytic carbon and graphite

International Nuclear Information System (INIS)

Matsuo, Hideto

1988-01-01

The effects of neutron irradiation and annealing on the temperature dependence of the linear thermal expansion of pyrolytic carbon and graphite were investigated after irradiation at 930-1280 0 C to a maximum neutron fluence of 2.84 x 10 25 m -2 (E > 29 fJ). After irradiation, little change in the thermal expansion of pyrolytic graphite was observed. However, as-deposited pyrolytic carbon showed an increase in thermal expansion in the perpendicular direction, a decrease in the direction parallel to the deposition plane, and also an increase in the anisotropy of the thermal expansion. Annealing at 2000 0 C did not cause any effective changes for irradiated specimens of either as-deposited pyrolytic carbon or pyrolytic graphite. (author)

Preparation and characterization of stearic acid/expanded graphite composites as thermal energy storage materials

International Nuclear Information System (INIS)

Fang, Guiyin; Li, Hui; Chen, Zhi; Liu, Xu

2010-01-01

Stearic acid/expanded graphite composites with different mass ratios were prepared by absorbing liquid stearic acid into the expanded graphite. In the composite materials, the stearic acid was used as the phase change material for thermal energy storage, and the expanded graphite acted as the supporting material. Fourier transformation infrared spectroscopy, X-ray diffraction, scanning electron microscopy and thermal diffusivity measurement were used to determine the chemical structure, crystalline phase, microstructure and thermal diffusivity of the composites, respectively. The thermal properties and thermal stability were investigated by differential scanning calorimetry and thermogravimetric analysis. The thermal analysis results indicated that the materials exhibited the same phase transition characteristics as the stearic acid and their latent heats were approximately the same as the values calculated based on the weight fraction of the stearic acid in the composites. The microstructural analysis results showed that the stearic acid was well absorbed in the porous network of the expanded graphite, and there was no leakage of the stearic acid from the composites even when it was in the molten state.

Porosity Effect on Thermal Properties of Al-12 wt % Si/Graphite Composites

Directory of Open Access Journals (Sweden)

José-Miguel Molina

2017-02-01

Full Text Available The effect of porosity on the thermal conductivity and the coefficient of thermal expansion of composites obtained by infiltration of Al-12 wt % Si alloy into graphite particulate preforms has been determined. Highly irregular graphite particles were used to fabricate the preforms. The thermal conductivity of these composites gradually increases with the applied infiltration pressure given the inherent reduction in porosity. A simple application of the Hasselman-Johnson model in a two-step procedure (that accounts for the presence of both graphite particles and voids randomly dispersed in a metallic matrix offers a good estimation of the experimental results. As concerns the coefficient of thermal expansion, the results show a slight increase with saturation being approximately in the range 14.6–15.2 × 10−6 K−1 for a saturation varying from 86% up to 100%. Results lie within the standard Hashin-Strikman bounds.

Porosity Effect on Thermal Properties of Al-12 wt % Si/Graphite Composites.

Science.gov (United States)

Molina, José-Miguel; Rodríguez-Guerrero, Alejandro; Louis, Enrique; Rodríguez-Reinoso, Francisco; Narciso, Javier

2017-02-14

The effect of porosity on the thermal conductivity and the coefficient of thermal expansion of composites obtained by infiltration of Al-12 wt % Si alloy into graphite particulate preforms has been determined. Highly irregular graphite particles were used to fabricate the preforms. The thermal conductivity of these composites gradually increases with the applied infiltration pressure given the inherent reduction in porosity. A simple application of the Hasselman-Johnson model in a two-step procedure (that accounts for the presence of both graphite particles and voids randomly dispersed in a metallic matrix) offers a good estimation of the experimental results. As concerns the coefficient of thermal expansion, the results show a slight increase with saturation being approximately in the range 14.6-15.2 × 10 -6 K -1 for a saturation varying from 86% up to 100%. Results lie within the standard Hashin-Strikman bounds.

Purification process of natural graphite as anode for Li-ion batteries: chemical versus thermal

Science.gov (United States)

Zaghib, K.; Song, X.; Guerfi, A.; Rioux, R.; Kinoshita, K.

The intercalation of Li ions in natural graphite that was purified by chemical and thermal processes was investigated. A new chemical process was developed that involved a mixed aqueous solution containing 30% H 2SO 4 and 30% NH xF y heated to 90 °C. The results of this process are compared to those obtained by heating the natural graphite from 1500 to 2400 °C in an inert environment (thermal process). The first-cycle coulombic efficiency of the purified natural graphite obtained by the chemical process is 91 and 84% after the thermal process at 2400 °C. Grinding the natural graphite before or after purification had no significant effect on electrochemical performance at low currents. However, grinding to a very small particle size before purification permitted optimization of the size distribution of the particles, which gives rise to a more homogenous electrode. The impurities in the graphite play a role as microabrasion agents during grinding which enhances its hardness and improves its mechanical properties. Grinding also modifies the particle morphology from a 2- to a 3-D structure (similar in shape to a potato). This potato-shaped natural graphite shows high reversible capacity at high current densities (about 90% at 1 C rate). Our analysis suggests that thermal processing is considerably more expensive than the chemical process to obtain purified natural graphite.

Graphite behaviour in relation to the fuel element design

Energy Technology Data Exchange (ETDEWEB)

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

1969-09-01

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

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)

Calculation of thermal stresses in graphite fuel blocks

International Nuclear Information System (INIS)

Lejeail, Y.; Cabrillat, M.T.

2005-01-01

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

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

International Nuclear Information System (INIS)

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

1975-01-01

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

Gas storage cylinder formed from a composition containing thermally exfoliated graphite

Science.gov (United States)

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

2012-01-01

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

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.

Characteristics of thermal neutron calibration fields using a graphite pile

International Nuclear Information System (INIS)

Uchita, Yoshiaki; Saegusa, Jun; Kajimoto, Yoichi; Tanimura, Yoshihiko; Shimizu, Shigeru; Yoshizawa, Michio

2005-03-01

The Facility of Radiation Standards of Japan Atomic Energy Research Institute is equipped with thermal neutron fields for calibrating area and personal neutron dosemeters. The fields use moderated neutrons leaked from a graphite pile in which radionuclide sources are placed. In January 2003, we have renewed the pile with some modifications in its size. In accordance with the renewal, we measured and calculated thermal neutron fluence rates, neutron energy distributions and angular distributions of the fields. The thermal neutron fluence rates of the ''inside-pile fields'' and the outside-pile fields'' were determined by the gold foil activation method. The neutron energy distributions of the outside-pile fields were also measured with the Bonner multi-sphere spectrometer system. The contributions of epithermal and fast neutrons to the total dose-equivalents were 9% in the southern outside-pile field and 12% in the western outside-pile field. The personal dose-equivalents, H p,slab (10, α), in the outside-pile fields are evaluated by considering the calculated angular distributions of incoming neutrons. The H p,slab (10, α) was found to be about 40% higher than the value in assuming the unidirectional neutron between the pile and the test point. (author)

Thermal shock fracture of graphite armor plate under the heat load of plasma disruption

International Nuclear Information System (INIS)

Horie, Tomoyoshi; Seki, Masahiro; Ohmori, Junji

1989-01-01

Experiments on the thermal shock brittle fracture of graphite plates were performed. Thermal loading which simulated a plasma disruption was produced by an electron beam facility. Pre-cracks produced on the surface propagated to the inside of the specimen even if the thermal stress on the surface was compressive. Two mechanisms are possible to produce tensile stress around the crack tip under thermal shock conditions. Temperature, thermal stress, and the stress intensity factor for the specimen were analyzed based on the finite element method for various heating conditions. The trend of experimental results under the asymmetric heating agrees qualitatively with the analytical results. This phenomenon is important for the design of plasma facing components made of graphite. Establishment of a lifetime prediction procedure including fatigue, fatigue crack growth, and brittle fracture is needed for graphite armors. (orig.)

Status of thermal neutron scattering data for graphite

International Nuclear Information System (INIS)

Mattes, M.; Keinert, J.

2005-07-01

At thermal neutron energies, the binding of the scattering nucleus in a solid, liquid, or gas affects the cross sections and the angular and energy distributions of the scattered neutrons. These effects are described in the thermal sub-library of evaluated files in File 7 of the ENDF-6 format. A re-evaluation of thermal neutron scattering data for carbon bound in graphite has been performed to investigate the impact of models (e.g., generalised frequency distributions) based on different experimental and theoretical data for the generation of scattering law data files S(α,β,T) and coherent elastic scattering data. Two phonon frequency distributions of graphite published in 2002 and 2004 were considered and the results compared with those based on the phonon spectra from Koppel et al. (published in 1968), on which the evaluations of ENDF/B-VI and JEFF-3.1 are based. The new frequency distributions were partly derived from ab initio simulations. Detailed comparisons with measurements of differential and integral neutron cross sections and other relevant data are reported. In addition, thermal MCNP data sets for use in the continuous Monte Carlo codes MCNP and MCNPX were generated from these evaluations for different temperatures. Calculated neutron spectra were found to be in good agreement with the measurements. (author)

New composites graphite/salt for high temperature thermal energy storage: From elaboration to development of thermal characterization methods for orthotropic conductive materials

International Nuclear Information System (INIS)

Acem, Zoubir

2007-01-01

This PhD is carried out within the framework of DISTOR (European) and HTPSTOCK (French) projects, which have for objective to conceive and study new graphite/salt composites dedicated to high temperature energy storage (>200 deg. C). She is split into two distinct part. The first one focused mainly on works linked with elaboration and thermal characterisation of these new composites. The different composites ways of elaboration (Dispersion, uniaxial compression, isostatic) associated to the different kind of graphite (Natural expanded graphite (ENG), synthetic graphite) investigated during the PhD are presented. The results of the thermal characterization campaign of these composites are also presented and permit to highlight the impact of graphite in the thermal behaviour of studied materials. Based on these results, modelling studies of the evolution of the thermal conductivity have been undertaken to deepen the understanding of the effect of graphite (quantity, size of particles) on the effective conductivity composites. The second one describes the thermal characterization devices and associated thermo-kinetics models which had to be developed and adapted to the specificities of newly developed materials. This concerns mainly the materials prepared by compression, which present orthotropic properties and are difficult to reproduce. So, the characterization of this kind of material is very difficult and tedious. That is why we are committed to develop and adapt existing methods of characterization to allow the complete thermal characterisation of an orthotropic conductive material from a single experimentation on a single sample. (author) [fr

Thermal conductivity degradation of graphites due to neutron irradiation at low temperature

International Nuclear Information System (INIS)

Snead, L.L.; Burchell, T.D.

1995-01-01

Several graphites and carbon/carbon composites (C/C's) have been irradiated with fission neutrons near 150 C and at fluences up to a displacement level of 0.24 dpa. The unirradiated room temperature thermal conductivity of these materials varied from 114 W/m K for H-451 isotropic graphite, to 670 W/m K for a unidirectional FMI-1D C/C composite. At the irradiation temperature a saturation reduction in thermal conductivity was seen to occur at displacement levels of approximately 0.1 dpa. All materials were seen to degrade to approximately 10 to 14% of their original thermal conductivity after irradiation. The significant recovery of thermal conductivity due to post-irradiation isochronal anneals is also presented. (orig.)

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.

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

Composites of aluminum alloy and magnesium alloy with graphite showing low thermal expansion and high specific thermal conductivity

Science.gov (United States)

Oddone, Valerio; Boerner, Benji; Reich, Stephanie

2017-12-01

High thermal conductivity, low thermal expansion and low density are three important features in novel materials for high performance electronics, mobile applications and aerospace. Spark plasma sintering was used to produce light metal-graphite composites with an excellent combination of these three properties. By adding up to 50 vol.% of macroscopic graphite flakes, the thermal expansion coefficient of magnesium and aluminum alloys was tuned down to zero or negative values, while the specific thermal conductivity was over four times higher than in copper. No degradation of the samples was observed after thermal stress tests and thermal cycling. Tensile strength and hardness measurements proved sufficient mechanical stability for most thermal management applications. For the production of the alloys, both prealloyed powders and elemental mixtures were used; the addition of trace elements to cope with the oxidation of the powders was studied.

Empirical Validation of Heat Transfer Performance Simulation of Graphite/PCM Concrete Materials for Thermally Activated Building System

Directory of Open Access Journals (Sweden)

Jin-Hee Song

2017-01-01

Full Text Available To increase the heat capacity in lightweight construction materials, a phase change material (PCM can be introduced to building elements. A thermally activated building system (TABS with graphite/PCM concrete hollow core slab is suggested as an energy-efficient technology to shift and reduce the peak thermal load in buildings. An evaluation of heat storage and dissipation characteristics of TABS in graphite/PCM concrete has been conducted using dynamic simulations, but empirical validation is necessary to acceptably predict the thermal behavior of graphite/PCM concrete. This study aimed to validate the thermal behavior of graphite/PCM concrete through a three-dimensional transient heat transfer simulation. The simulation results were compared to experimental results from previous studies of concrete and graphite/PCM concrete. The overall thermal behavior for both materials was found to be similar to experiment results. Limitations in the simulation modeling, which included determination of the indoor heat transfer coefficient, assumption of constant thermal conductivity with temperature, and assumption of specimen homogeneity, led to slight differences between the measured and simulated results.

THERMAL INSULATION PROPERTIES RESEARCH OF THE COMPOSITE MATERIAL WATER GLASS–GRAPHITE MICROPARTICLES

Directory of Open Access Journals (Sweden)

V. A. Gostev

2014-05-01

Full Text Available Research results for the composite material (CM water glass–graphite microparticles with high thermal stability and thermal insulation properties are given. A composition consisting of graphite (42 % by weight, water glass Na2O(SiO2n (50% by weight and the hardener - sodium silicofluoric Na2SiF6 (8% by weight. Technology of such composition receipt is suggested. Experimental samples of the CM with filler particles (graphite and a few microns in size were obtained. This is confirmed by a study of samples by X-ray diffraction and electron microscopy. The qualitative and quantitative phase analysis of the CM structure is done. Load limit values leading to the destruction of CM are identified. The character of the rupture surface is detected. Numerical values of specific heat and thermal conductivity are defined. Dependence of the specific heat capacity and thermal conductivity on temperature at monotonic heating is obtained experimentally. Studies have confirmed the increased thermal insulation properties of the proposed composition. CM with such characteristics can be recommended as a coating designed to reduce heat losses and resistant to high temperatures. Due to accessibility and low cost of its components the proposed material can be produced on an industrial scale.

Graphite Foam Heat Exchangers for Thermal Management

Energy Technology Data Exchange (ETDEWEB)

Klett, J.W.

2004-06-07

Improved thermal management is needed to increase the power density of electronic and more effectively cool electronic enclosures that are envisioned in future aircraft, spacecraft and surface ships. Typically, heat exchanger cores must increase in size to more effectively dissipate increased heat loads, this would be impossible in many cases, thus improved heat exchanger cores will be required. In this Phase I investigation, MRi aimed to demonstrate improved thermal management using graphite foam (Gr-foam) core heat exchangers. The proposed design was to combine Gr-foams from POCO with MRi's innovative low temperature, active metal joining process (S-Bond{trademark}) to bond Gr-foam to aluminum, copper and aluminum/SiC composite faceplates. The results were very favorable, so a Phase II SBIR with the MDA was initiated. This had primarily 5 tasks: (1) bonding, (2) thermal modeling, (3) cooling chip scale packages, (4) evaporative cooling techniques and (5) IGBT cold plate development. The bonding tests showed that the ''reflow'' technique with S-Bond{reg_sign}-220 resulted in the best and most consistent bond. Then, thermal modeling was used to design different chip scale packages and IGBT cold plates. These designs were used to fabricate many finned graphite foam heat sinks specifically for two standard type IC packages, the 423 and 478 pin chips. These results demonstrated several advantages with the foam. First, the heat sinks with the foam were lighter than the copper/aluminum sinks used as standards. The sinks for the 423 design made from foam were not as good as the standard sinks. However, the sinks made from foam for the 478 pin chips were better than the standard heat sinks used today. However, this improvement was marginal (in the 10-20% better regime). However, another important note was that the epoxy bonding technique resulted in heat sinks with similar results as that with the S-bond{reg_sign}, slightly worse than the S

Improvement of thermal shock resistance of isotropic graphite by ti-doping

International Nuclear Information System (INIS)

Lopez-Galilea, I.; Ordas, N.; Garcia-Rosales, C.; Lindig, S.

2007-01-01

Full text of publication follows: Carbon fiber reinforced carbon (CFC) is the present candidate material for the strike point area of the ITER divertor due to its ability to withstand excessive heat loads during ELMs and plasma disruptions. However, chemical erosion of carbon under hydrogen bombardment from the plasma involves serious disadvantages for this application (replacement and safety problems due to tritium co-deposition). In addition, the manufacturing process of present CFC candidate materials is long and complex resulting in high costs, and CFC materials are inherently anisotropic. Doping of carbon with small amounts (several at. %) of titanium has proved to be effective in reducing chemical erosion while maintaining or even improving the mechanical properties. furthermore, TiC as dopant contributes to increase significantly the thermal conductivity and consequently the thermal shock resistance, due to the catalytic effect of this carbide on the graphitization. The aim of this work is to improve substantially the thermal shock resistance of fine-grained isotropic graphite by doping it with small amounts of TiC, reducing at the same time the chemical erosion. By this way Ti-doped graphites could be competitive with present CFC candidate materials for next step fusion devices. To achieve this, a synthetic naphthalene-derived mesophase pitch named AR is used as carbon precursor; this raw material exhibits excellent graphitizability, high chemical purity and consistent quality. Due to the low viscosity at the softening point of AR, resulting in swelling during the carbonization treatment, it is necessary to modify the initial viscosity of AR by an adequate oxidative stabilization treatment. As dopant, TiC powder with 130 nm average particle size is added. The influence of several manufacturing parameters such as oxidative stabilization treatment, carbonization cycle, graphitization temperature and dwell time during graphitization have been investigated in

Improvement of thermal shock resistance of isotropic graphite by ti-doping

Energy Technology Data Exchange (ETDEWEB)

Lopez-Galilea, I.; Ordas, N.; Garcia-Rosales, C. [Navarrra Univ., CEPT, San Sebastian (Spain); Lindig, S. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany)

2007-07-01

Full text of publication follows: Carbon fiber reinforced carbon (CFC) is the present candidate material for the strike point area of the ITER divertor due to its ability to withstand excessive heat loads during ELMs and plasma disruptions. However, chemical erosion of carbon under hydrogen bombardment from the plasma involves serious disadvantages for this application (replacement and safety problems due to tritium co-deposition). In addition, the manufacturing process of present CFC candidate materials is long and complex resulting in high costs, and CFC materials are inherently anisotropic. Doping of carbon with small amounts (several at. %) of titanium has proved to be effective in reducing chemical erosion while maintaining or even improving the mechanical properties. furthermore, TiC as dopant contributes to increase significantly the thermal conductivity and consequently the thermal shock resistance, due to the catalytic effect of this carbide on the graphitization. The aim of this work is to improve substantially the thermal shock resistance of fine-grained isotropic graphite by doping it with small amounts of TiC, reducing at the same time the chemical erosion. By this way Ti-doped graphites could be competitive with present CFC candidate materials for next step fusion devices. To achieve this, a synthetic naphthalene-derived mesophase pitch named AR is used as carbon precursor; this raw material exhibits excellent graphitizability, high chemical purity and consistent quality. Due to the low viscosity at the softening point of AR, resulting in swelling during the carbonization treatment, it is necessary to modify the initial viscosity of AR by an adequate oxidative stabilization treatment. As dopant, TiC powder with 130 nm average particle size is added. The influence of several manufacturing parameters such as oxidative stabilization treatment, carbonization cycle, graphitization temperature and dwell time during graphitization have been investigated in

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

Energy Technology Data Exchange (ETDEWEB)

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

1948-11-01

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

Thermal deoxygenation of graphite oxide at low temperature

International Nuclear Information System (INIS)

Kampars, V; Legzdina, M

2015-01-01

Synthesis of graphene via the deoxygenation of the graphite oxide (GO) is a method for the large-scale production of this nanomaterial possessing exceptional mechanical, electrical and translucent properties. Graphite oxide sheet contains at least four different oxygen atoms connected to the Csp 3 and Csp 2 atoms of the sheet in the form of hydroxyl, epoxy, carboxyl or carbonyl groups. Some of these functional groups are located at the surface but others situated at the edges of the platelets. To obtain the graphene nanoplatelets or the few-layer graphene the oxygen functionalities must be removed. Exfoliation and deoxygenation can be accomplished by the use of chemical reductants or heat. Thermal deoxygenation as greener and simpler approach is more preferable over chemical reduction approach. Usually a considerable mass loss of GO observed upon heating at temperatures starting at 200 °C and is attributed to the deoxygenation process. In order to avoid the defects of the obtained graphene sheets it is very important to find the methods for lowering the deoxygenation temperature of GO. Herein, we have investigated the way treatment of the Hummer's synthesis product with acetone and methyl tert-butyl ether under ultrasonication in order to lower the thermal stability of the graphite oxide and its deoxygenation temperature. The obtained results indicate that treatment of the graphite oxide with solvents mentioned above substantially reduces the reduction and exfoliation temperature (130 °C) under ambient atmosphere. The investigation of the composition of evolved gases by hyphenated Pyr/GC/MS method at different experimental conditions under helium atmosphere shows that without the expected H 2 O, CO and CO 2 also sulphur dioxide and acetone has been released

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

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)

Thermal oxidation of nuclear graphite: A large scale waste treatment option

Science.gov (United States)

Jones, Abbie N.; Marsden, Barry J.

2017-01-01

This study has investigated the laboratory scale thermal oxidation of nuclear graphite, as a proof-of-concept for the treatment and decommissioning of reactor cores on a larger industrial scale. If showed to be effective, this technology could have promising international significance with a considerable impact on the nuclear waste management problem currently facing many countries worldwide. The use of thermal treatment of such graphite waste is seen as advantageous since it will decouple the need for an operational Geological Disposal Facility (GDF). Particulate samples of Magnox Reactor Pile Grade-A (PGA) graphite, were oxidised in both air and 60% O2, over the temperature range 400–1200°C. Oxidation rates were found to increase with temperature, with a particular rise between 700–800°C, suggesting a change in oxidation mechanism. A second increase in oxidation rate was observed between 1000–1200°C and was found to correspond to a large increase in the CO/CO2 ratio, as confirmed through gas analysis. Increasing the oxidant flow rate gave a linear increase in oxidation rate, up to a certain point, and maximum rates of 23.3 and 69.6 mg / min for air and 60% O2 respectively were achieved at a flow of 250 ml / min and temperature of 1000°C. These promising results show that large-scale thermal treatment could be a potential option for the decommissioning of graphite cores, although the design of the plant would need careful consideration in order to achieve optimum efficiency and throughput. PMID:28793326

Thermal oxidation of nuclear graphite: A large scale waste treatment option.

Directory of Open Access Journals (Sweden)

Alex Theodosiou

Full Text Available This study has investigated the laboratory scale thermal oxidation of nuclear graphite, as a proof-of-concept for the treatment and decommissioning of reactor cores on a larger industrial scale. If showed to be effective, this technology could have promising international significance with a considerable impact on the nuclear waste management problem currently facing many countries worldwide. The use of thermal treatment of such graphite waste is seen as advantageous since it will decouple the need for an operational Geological Disposal Facility (GDF. Particulate samples of Magnox Reactor Pile Grade-A (PGA graphite, were oxidised in both air and 60% O2, over the temperature range 400-1200°C. Oxidation rates were found to increase with temperature, with a particular rise between 700-800°C, suggesting a change in oxidation mechanism. A second increase in oxidation rate was observed between 1000-1200°C and was found to correspond to a large increase in the CO/CO2 ratio, as confirmed through gas analysis. Increasing the oxidant flow rate gave a linear increase in oxidation rate, up to a certain point, and maximum rates of 23.3 and 69.6 mg / min for air and 60% O2 respectively were achieved at a flow of 250 ml / min and temperature of 1000°C. These promising results show that large-scale thermal treatment could be a potential option for the decommissioning of graphite cores, although the design of the plant would need careful consideration in order to achieve optimum efficiency and throughput.

TAPIR, Thermal Analysis of HTGR with Graphite Sleeve Fuel Elements

International Nuclear Information System (INIS)

Weicht, U.; Mueller, W.

1983-01-01

1 - Nature of the physical problem solved: Thermal analysis of a reactor core containing internally and/or externally gas cooled prismatic fuel elements of various geometries, rating, power distribution, and material properties. 2 - Method of solution: A fuel element in this programme is regarded as a sector of a fuelled annulus with graphite sleeves of any shape on either side and optional annular gaps between fuel and graphite and/or within the graphite. It may have any centre angle and the fuelled annulus may become a solid cylindrical rod. Heat generation in the fuel is assumed to be uniform over the cross section and peripheral heat flux into adjacent sectors is ignored. Fuel elements and coolant channels are treated separately, then linked together to fit a specified pattern. 3 - Restrictions on the complexity of the problem: Maxima of: 50 fuel elements; 50 cooled channels; 25 fuel geometries; 25 coolant channel geometries; 10 axial power distributions; 10 graphite conductivities

Evaluation of thermal shock strengths for graphite materials using a laser irradiation method

International Nuclear Information System (INIS)

Kim, Jae Hoon; Lee, Young Shin; Kim, Duck Hoi; Park, No Seok; Suh, Jeong; Kim, Jeng O.; Il Moon, Soon

2004-01-01

Thermal shock is a physical phenomenon that occurs during the exposure to rapidly high temperature and pressure changes or during quenching of a material. The rocket nozzle throat is exposed to combustion gas of high temperature. Therefore, it is important to select suitable materials having the appropriate thermal shock resistance and to evaluate these materials for rocket nozzle design. The material of this study is ATJ graphite, which is the candidate material for rocket nozzle throat. This study presents an experimental method to evaluate the thermal shock resistance and thermal shock fracture toughness of ATJ graphite using laser irradiation. In particular, thermal shock resistance tests are conducted with changes of specimen thickness, with laser source irradiated at the center of the specimen. Temperature distributions on the specimen surface are detected using type K and C thermocouples. Scanning electron microscope (SEM) is used to observe the thermal cracks on specimen surface

An analysis of irradiation creep in nuclear graphites

International Nuclear Information System (INIS)

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

2002-01-01

Nuclear graphite under load shows remarkably high creep ductility with neutron irradiation, well in excess of any strain experienced in un-irradiated graphite (and additional to any dimensional changes that would occur without stress). As this behaviour compensates, to some extent, some other irradiation effects such as thermal shutdown stresses, it is an important property. This paper briefly reviews the approach to irradiation creep in the UK, described by the UK Creep Law. It then offers an alternative analysis of irradiation creep applicable to most situations, including HTR systems, using AGR moderator graphite as an example, to high values of neutron fluence, applied stress and radiolytic weight loss. (authors)

Magnetic susceptibilities and thermal expansion of artificial graphites

International Nuclear Information System (INIS)

Cornuault, P.; Herpin, A.; Hering, H.; Seguin, M.; Commissariat a l'Energie Atomique, Saclay

1960-01-01

Starting from measurements of the magnetic susceptibility made in the two principal directions of a graphite bar, the distribution function of the normals to the carbon planes in the crystallites has been evaluated. The effect of different variation in the manufacturing process on this crystalline anisotropy has been studied. From this crystalline anisotropy we have calculated the thermal expansion coefficient possessed by a compact mass of crystallites having exactly the same orientational anisotropy as the porous body consideration. The difference between this and the observed expansion coefficient leads to the determination of the expansion of the non-graphitic part of the mass which turns out to have a negative value and is also anisotropic. We have attempted to draw some conclusions from this result. (author) [fr

Electrospray painted article containing thermally exfoliated graphite oxide and method for their manufacture

Science.gov (United States)

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

2011-01-01

A painted polymer part containing a conductive polymer composition containing at least one polymer and a modified graphite oxide material, which is a thermally exfoliated graphite oxide with a surface area of from about 300 sq m/g to 2600 sq m/g, wherein the painted polymer part has been electrospray painted.

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)

Experimental investigation of linear thermal expansion of pyrolytic graphite at high temperatures

Science.gov (United States)

Senchenko, V. N.; Belikov, R. S.

2017-11-01

Using the previously described [1] experimental setup for investigation of the thermophysical properties of refractory materials under high pressure and temperature a few experiments with pyrolytic graphite were carried out. The density of the material was equal to 2.18 g/cm3. Experimental data on the linear thermal expansion in the perpendicular and parallel to the basal plane direction were obtained. Thermal expansion in the perpendicular to the basal plane direction during the heating from room temperature up to the melting point was 16.4 ± 1.6%. The results obtained allow calculating the density of pyrolytic graphite in the wide range of high temperatures up to the melting point.

The moderator's moderator

International Nuclear Information System (INIS)

Williamson, G.K.

1990-01-01

A brief account is given of the development of graphite moderators for Magnox and advanced gas cooled reactors. The accident at Windscale in 1957 brought to worldwide attention the importance of irradiation damage in graphite and the consequent storage of Wigner energy. In spite of the Windscale setback, preparations for the civil programme of Magnox reactors went ahead apace. Some of the background to the disastrous Dungeness B tender is presented. In spite of all the difficulties and uncertainties, the graphite in UK reactors has performed well. In all cases, as far as the author is aware, the behaviour of the graphite moderators will not prevent design life being achieved. (author)

Studies on thermal neutron perturbation factor needed for bulk sample activation analysis

CERN Document Server

Csikai, J; Sanami, T; Michikawa, T

2002-01-01

The spatial distribution of thermal neutrons produced by an Am-Be source in a graphite pile was measured via the activation foil method. The results obtained agree well with calculated data using the MCNP-4B code. A previous method used for the determination of the average neutron flux within thin absorbing samples has been improved and extended for a graphite moderator. A procedure developed for the determination of the flux perturbation factor renders the thermal neutron activation analysis of bulky samples of unknown composition possible both in hydrogenous and graphite moderators.

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Development of graphite foam infiltrated with MgCl 2 for a latent heat based thermal energy storage (LHTES) system

Energy Technology Data Exchange (ETDEWEB)

Singh, Dileep; Kim, Taeil; Zhao, Weihuan; Yu, Wenhua; France, David M.

2016-08-01

Thermal energy storage (TES) systems that are compatible with high temperature power cycles for concentrating solar power (CSP) require high temperature media for transporting and storing thermal energy. To that end, TES systems have been proposed based on the latent heat of fusion of the phase change materials (PCMs). However, PCMs have relatively low thermal conductivities. In this paper, use of high-thermal-conductivity graphite foam infiltrated with a PCM (MgCl2) has been investigated as a potential TES system. Graphite foams with two porosities were infiltrated with MgCl2. The infiltrated composites were evaluated for density, heat of fusion, melting/freezing temperatures, and thermal diffusivities. Estimated thermal conductivities of MgCl2/graphite foam composites were significantly higher than those of MgCl2 alone over the measured temperature range. Furthermore, heat of fusion, melting/freezing temperatures, and densities showed comparable values to those of pure MgCl2. Results of this study indicate that MgCl2/graphite foam composites show promise as storage media for a latent heat thermal energy storage system for CSP applications.

Thermal cyclic oxidation behavior of the developed compositionally gradient graphite material of SiC/C in air environment

International Nuclear Information System (INIS)

Nakano, Junichi; Fujii, Kimio; Shindo, Masami

1993-08-01

For the developed compositionally gradient graphite material composed of surface SiC coating layer, middle SiC/C layer and graphite matrix, the thermal cyclic oxidation test was performed together with two kinds of the SiC coated graphite materials in air environment. It was made clear that the developed material exhibited high performance under severe thermal cyclic condition independent of the morphology of middle SiC/C layers and had the longer time or the more cycle margins from crack initiation to failure for surface SiC coating layer compared with the SiC coated graphite materials. (author)

A core-monitoring based methodology for predictions of graphite weight loss in AGR moderator bricks

Energy Technology Data Exchange (ETDEWEB)

McNally, K., E-mail: kevin.mcnally@hsl.gsi.gov.uk [Health and Safety Laboratory, Harpur Hill, Buxton, Derbyshire SK17 9JN (United Kingdom); Warren, N. [Health and Safety Laboratory, Harpur Hill, Buxton, Derbyshire SK17 9JN (United Kingdom); Fahad, M.; Hall, G.; Marsden, B.J. [Nuclear Graphite Research Group, School of MACE, University of Manchester, Manchester M13 9PL (United Kingdom)

2017-04-01

Highlights: • A statistically-based methodology for estimating graphite density is presented. • Graphite shrinkage is accounted for using a finite element model. • Differences in weight loss forecasts were found when compared to the existing model. - Abstract: Physically based models, resolved using the finite element (FE) method are often used to model changes in dimensions and the associated stress fields of graphite moderator bricks within a reactor. These models require inputs that describe the loading conditions (temperature, fluence and weight loss ‘field variables’), and coded relationships describing the behaviour of graphite under these conditions. The weight loss field variables are calculated using a reactor chemistry/physics code FEAT DIFFUSE. In this work the authors consider an alternative data source of weight loss: that from a longitudinal dataset of density measurements made on small samples trepanned from operating reactors during statutory outages. A nonlinear mixed-effect model is presented for modelling the age and depth-related trends in density. A correction that accounts for irradiation-induced dimensional changes (axial and radial shrinkage) is subsequently applied. The authors compare weight loss forecasts made using FEAT DIFFUSE with those based on an alternative statistical model for a layer four moderator brick for the Hinkley Point B, Reactor 3. The authors compare the two approaches for the weight loss distribution through the brick with a particular focus on the interstitial keyway, and for the average (over the volume of the brick) weight loss.

Thermal performance measurements of a graphite tube compact cryogenic support for the Superconducting Super Collider

International Nuclear Information System (INIS)

Gonczy, J.D.; Boroski, W.N.; Larson, E.T.; Nicol, T.H.; Niemann, R.C.; Otavka, J.G.; Ruschman, M.K.

1988-12-01

The magnet cryostat development program for the Superconducting Super Collider (SSC) High Energy Physics Proton-Proton Collider has produced an innovative design for the structural support of the cold mass and thermal radiation shields. This work updates the continuing development of the support known as the Compact Cryogenic Support (CCS). As the structural and thermal requirements of the SSC became better defined, a CCS was developed that employs an innermost tube comprised of a graphite composite material. Presented is the thermal performance to 4.5K of the graphite CCS model. 8 refs., 6 figs., 2 tabs

High thermal conductivity of graphite fiber silicon carbide composites for fusion reactor application

International Nuclear Information System (INIS)

Snead, L.L.; Balden, M.; Causey, R.A.; Atsumi, H.

2002-01-01

The benefits of using CVI SiC/graphite fiber composites as low tritium retaining, high thermal conductivity composites for fusion applications are presented. Three-dimensional woven composites have been chemically vapor infiltrated with SiC and their thermophysical properties measured. One material used an intermediate grade graphite fiber in all directions (Amoco P55) while a second material used very high thermal conductive fiber (Amoco K-1100) in the high fiber density direction. The overall void was less than 20%. Strength as measured by four-point bending was comparable to those of SiC/SiC composite. The room temperature thermal conductivity in the high conductivity direction was impressive for both materials, with values >70 W/m K for the P-55 and >420 W/m K for the K-1100 variant. The thermal conductivity was measured as a function of temperature and exceeds the highest thermal conductivity of CVD SiC currently available at fusion relevant temperatures (>600 deg. C). Limited data on the irradiation-induced degradation in thermal conductivity is consistent with carbon fiber composite literature

High Thermal Conductivity of Copper Matrix Composite Coatings with Highly-Aligned Graphite Nanoplatelets

Science.gov (United States)

Tagliaferri, Vincenzo; Ucciardello, Nadia

2017-01-01

Nanocomposite coatings with highly-aligned graphite nanoplatelets in a copper matrix were successfully fabricated by electrodeposition. For the first time, the disposition and thermal conductivity of the nanofiller has been evaluated. The degree of alignment and inclination of the filling materials has been quantitatively evaluated by polarized micro-Raman spectroscopy. The room temperature values of the thermal conductivity were extracted for the graphite nanoplatelets by the dependence of the Raman G-peak frequency on the laser power excitation. Temperature dependency of the G-peak shift has been also measured. Most remarkable is the global thermal conductivity of 640 ± 20 W·m−1·K−1 (+57% of copper) obtained for the composite coating by the flash method. Our experimental results are accounted for by an effective medium approximation (EMA) model that considers the influence of filler geometry, orientation, and thermal conductivity inside a copper matrix. PMID:29068424

Continuous Carbon Nanotube-Ultrathin Graphite Hybrid Foams for Increased Thermal Conductivity and Suppressed Subcooling in Composite Phase Change Materials.

Science.gov (United States)

Kholmanov, Iskandar; Kim, Jaehyun; Ou, Eric; Ruoff, Rodney S; Shi, Li

2015-12-22

Continuous ultrathin graphite foams (UGFs) have been actively researched recently to obtain composite materials with increased thermal conductivities. However, the large pore size of these graphitic foams has resulted in large thermal resistance values for heat conduction from inside the pore to the high thermal conductivity graphitic struts. Here, we demonstrate that the effective thermal conductivity of these UGF composites can be increased further by growing long CNT networks directly from the graphite struts of UGFs into the pore space. When erythritol, a phase change material for thermal energy storage, is used to fill the pores of UGF-CNT hybrids, the thermal conductivity of the UGF-CNT/erythritol composite was found to increase by as much as a factor of 1.8 compared to that of a UGF/erythritol composite, whereas breaking the UGF-CNT bonding in the hybrid composite resulted in a drop in the effective room-temperature thermal conductivity from about 4.1 ± 0.3 W m(-1) K(-1) to about 2.9 ± 0.2 W m(-1) K(-1) for the same UGF and CNT loadings of about 1.8 and 0.8 wt %, respectively. Moreover, we discovered that the hybrid structure strongly suppresses subcooling of erythritol due to the heterogeneous nucleation of erythritol at interfaces with the graphitic structures.

Neutronic analysis of graphite-moderated solid breeder design for INTOR

International Nuclear Information System (INIS)

Jung, J.; Abdou, M.A.

1981-01-01

An in-depth analysis of the INTOR tritium-production-blanket design is presented. A ternary system of solid silicate breeder, lead neutron multiplier, and graphite moderator is explored primary from safety and blanket tritium-inventory considerations. Lithium-silicate (Li 2 SiO 3 ) breeder systems are studied along with water (H 2 O/D 2 O) and Type 316 stainless steel as coolant and structural material, respectively. The analysis examines the neutronics effects on tritium-production regarding: (1) coolant choice; (2) moderator choice; (3) moderator location; (4) multiplier thickness; (5) 6 Li enrichment; and (6) 6 Li burnup. The tritium-breeding-blanket modules are located at the top, outboard, and bottom (outer) parts of the torus, resulting in a breeding coverage of approx. 60% at the first-wall surface. It is found that the reference INTOR design yields, based on a three-dimensional analysis, a net tritium breeding ratio (BR) of approx. 0.65 at the beginning of reactor operation, satisfying the design criterion of BR > 0.6

Preparation of an aqueous graphitic ink for thermal drop-on-demand inkjet printing

Energy Technology Data Exchange (ETDEWEB)

Romagnoli, Marcello; Lassinantti Gualtieri, Magdalena, E-mail: magdalena.gualtieri@unimore.it; Cannio, Maria; Barbieri, Francesco; Giovanardi, Roberto

2016-10-01

A graphitic ink for thermal DOD inkjet printing was developed. Challenges to be met were related to the small size of the getting nozzle (20 μm), demanding high dispersion stability of submicron particles, as well as to the physical requirements of the printer. In addition, solvents potentially hazardous to human health were excluded a priori. These necessities led to the development of a ternary aqueous solvent system based on 2-propanol and monoethylene glycol, offering an environmental-friendly alternative to conventional graphene solvents. In addition, high flexibility in terms of physical properties (e.g. surface tension, viscosity, density) important for jetting is obtained. Size reduction and exfoliation, accomplished by wet-grinding of graphite in the presence of a surfactant, were followed by laser diffraction and XRD line broadening analyses, respectively. The separated graphitic colloids used for preparation of inks were composed of ca 30 layers of AB–stacked graphene flakes, as determined by line broadening analyses (XRD data). Jetting of an ink with a solid content of 0.3 mg/mL gave a thickness increase of ca. 25 nm/pass, as determined by FESEM. Electrical characterization evidenced the need to remove residual organic molecules to regain the electrical properties of the graphitic particles. - Highlights: • A non-hazardous graphitic ink for thermal DOD inkjet printing was developed. • The ternary mixture water/ethylene glycol/2-propanol is suitable as solvent. • Physical properties important for jetting is tailored by solvent composition. • Surfactant-aided grinding gives exfoliation of graphite without inflicting microstrain.

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)

Wear Behavior of Selected Nuclear Grade Graphites at Room Temperature in Ambient Air Environment

International Nuclear Information System (INIS)

Kim, Eung-Seon; Park, Kwang-Seok; Kim, Yong-Wan

2008-01-01

In a very high temperature reactor (VHTR), graphite will be used not only for as a moderator and reflector but also as a major structural component due to its excellent neutronic, thermal and mechanical properties. In the VHTR, wear of graphite components is inevitable due to a neutron irradiation-induced dimensional change, thermal gradient, relative motions of graphite components and a shock load such as an earthquake. Large wear particles accumulated at the bottom of a reactor can influence the cooling of the lower part and small wear particles accumulated on the primary circuit and heat exchanger tube can make it difficult to inspect the equipment, and also decrease the heat exchange rate. In the present work, preliminary wear tests were performed at room temperature in ambient air environment to understand the basic wear characteristics of selected nuclear grade graphites for the VHTR

Studies of physicochemical properties of graphite oxide and thermally exfoliated/reduced graphene oxide

Directory of Open Access Journals (Sweden)

Drewniak Sabina Elżbieta

2015-12-01

Full Text Available The aim of the experimental research studies was to determine some electrical properties of graphite oxide and thermally exfoliated/reduced graphene oxide. The authors tried to interpret the obtained physicochemical results. For that purpose, both resistance measurements and investigation studies were carried out in order to characterize the samples. The resistance was measured at various temperatures in the course of composition changes of gas atmospheres (which surround the samples. The studies were also supported by such methods as: scanning electron microscopy (SEM, Raman spectroscopy (RS, atomic force microscopy (AFM and thermogravimetry (TG. Moreover, during the experiments also the elemental analyses (EA of the tested samples (graphite oxide and thermally exfoliated/reduced graphene oxide were performed.

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

International Nuclear Information System (INIS)

Baltazar, O.

1978-01-01

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

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.)

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

Design of a graphite-moderated {sup 241}Am-Li neutron field to simulate reactor spectra

Energy Technology Data Exchange (ETDEWEB)

Tsujimura, N., E-mail: tsujimura.norio@jaea.go.j [Nuclear Fuel Cycle Engineering Laboratories, Japan Atomic Energy Agency, 4-33, Tokai-mura, Ibaraki-ken, 319-1194 (Japan); Yoshida, T. [Nuclear Fuel Cycle Engineering Laboratories, Japan Atomic Energy Agency, 4-33, Tokai-mura, Ibaraki-ken, 319-1194 (Japan)

2010-12-15

A neutron calibration field using {sup 241}Am-Li sources and a moderator was designed to simulate the neutron fields found outside a reactor. The moderating assembly selected for the design calculation consists of a cube of graphite blocks with dimensions of 50 cm by 50 cm by 50 cm, in which the {sup 241}Am-Li sources are placed. Monte Carlo calculations revealed the optimal depth of the source to be 15 cm. This moderated neutron source can be used to provide a test field that has a large number of intermediate energy neutrons with a small portion of MeV component.

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.

A nano-graphite/paraffin phase change material with high thermal conductivity

International Nuclear Information System (INIS)

Li, Min

2013-01-01

Highlights: ► Paraffin and NG formed a nanoscale compound. ► The thermal conductivity increased gradually with the content of NG. ► The thermal conductivity of the material containing 10% NG were 0.9362 W/m K. - Abstract: Nano-graphite (NG)/paraffin composites were prepared as composite phase change materials. NG has the function of improving the thermal conductivity of the composite. The microstructure and thermal properties of the materials were examined with environmental scanning electron microscopy and differential scanning calorimetry. The results indicated that the NG layers were randomly dispersed in the paraffin, and the thermal conductivity increased gradually with the content of NG. Thermal conductivity of the material containing 10% NG were 0.9362 W/m K

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

International Nuclear Information System (INIS)

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

1992-02-01

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

Thermal decomposition of cesium-ethylene-ternary graphite intercalation compounds

International Nuclear Information System (INIS)

Matsumoto, R.; Oishi, Y.; Arii, T.

2010-01-01

In this paper, the thermal decomposition of air-stable Cs-ethylene-ternary graphite intercalation compounds (GICs) is discussed. The air stability of Cs-GICs is improved remarkably after the absorption of ethylene into their interlayer nanospace, because the ethylene molecules oligomerize and block the movement of Cs atoms. In addition, the evaporation of Cs atoms from the Cs-ethylene-ternary GICs is observed above 400 o C under a N 2 atmosphere of 100 Pa by ion attachment mass spectrometry. Although the results indicate that Cs-ethylene-ternary GICs remain stable up to approximately 400 o C, their thermal stability is not very high as compared to that of Cs-GICs.

Applications of graphite-enabled phase change material composites to improve thermal performance of cementitious materials

Science.gov (United States)

Li, Mingli; Lin, Zhibin; Wu, Lili; Wang, Jinhui; Gong, Na

2017-11-01

Enhancing the thermal efficiency to decrease the energy consumption of structures has been the topic of much research. In this study, a graphite-enabled microencapsulated phase change material (GE-MEPCM) was used in the production of a novel thermal energy storage engineered cementitious composite feathering high heat storage capacity and enhanced thermal conductivity. The surface morphology and particle size of the microencapsulated phase change material (MEPCM) were investigated by scanning electron microscopy (SEM). Thermal properties of MEPCM was determined using differential scanning calorimetry (DSC). In addition, thermal and mechanical properties of the cementitious mortar with different admixtures were explored and compared with those of a cementitious composite. It was shown that the latent heat of MEPCM was 162 J/g, offering much better thermal energy storage capacity to the cementitious composite. However, MEPCM was found to decrease the thermal conductivity of the composite, which can be effectively solved by adding natural graphite (NG). Moreover, the incorporation of MEPCM has a certain decrease in the compressive strength, mainly due to the weak interfaces between MEPCM and cement matrix.

Pulsed neutron intensity from rectangular shaped light water moderator with fast-neutron reflector

International Nuclear Information System (INIS)

Kiyanagi, Yoshiaki; Iwasa, Hirokatsu

1982-01-01

With a view to enhancing the thermal-neutron intensity obtained from a pulsed neutron source, an experimental study has been made to determine the optimum size of a rectangular shaped light water moderator provided with fast neutron reflector of beryllium oxide or graphite, and decoupled thermal-neutronically by means of Cd sheet. The optimum dimensions for the moderator are derived for the neutron emission surface and the thickn ess, for the cases in which the neutron-producing target is placed beneath the moderator (''wing geometry'') or immediately behind the moderator (''slab geometry''). The major conclusions drawn from the experimental results are as follows. The presence of the Cd decoupler inserted between the moderator and reflector prevent the enhancement of thermal-neutron emission time gained by the provision of reflector. With a graphite reflector about 14 cm thick, (a) the optimum area of emission surface would be 25 x 25 cm 2 for wing geometry and still larger for slab geometry, and (b) the optimum moderator thickness would be 5.5 cm for slab geometry and 8.5 cm for wing geometry. It is thus concluded that a higher neutron emission intensity can be obtained with slab than with wing geometry provided that a large emission surface can be adopted for the moderator. (author)

Thermal gradients caused by the CANDU moderator circulation

International Nuclear Information System (INIS)

Mohindra, V.K.; Vartolomei, M.A.; Scharfenberg, R.

2008-01-01

The heavy water moderator circulation system of a CANDU reactor, maintains calandria moderator temperature at power-dependent design values. The temperature differentials between the moderator and the cooler heavy water entering the calandria generate thermal gradients in the reflector and moderator. The resultant small changes in thermal neutron population are detected by the out-of-core ion chambers as small, continuous fluctuations of the Log Rate signals. The impact of the thermal gradients on the frequency of the High Log Rate fluctuations and their amplitude is relatively more pronounced for Bruce A as compared to Bruce B reactors. The root cause of the Log Rate fluctuations was investigated using Bruce Power operating plant information data and the results of the investigation support the interpretation based on the thermal gradient phenomenon. (author)

Modification of graphite structure by irradiation, revealed by thermal oxidation. Examination by electronic microscopy

International Nuclear Information System (INIS)

Rouaud, Michel

1969-01-01

Based on the analysis of images obtained by electronic microscopy, this document reports the comparative study of the action of neutrons on three different graphites: a natural one (Ticonderoga) and two pyrolytic ones (Carbone-Lorraine and Raytheon). The approach is based on the modification of features of thermal oxidation of graphites by dry air after irradiation. Different corrosion features are identified. The author states that there seems to be a relationship between the number and shape of these features, and defects existing on the irradiated graphite before oxidation. For low doses, the feature aspect varies with depth at which oxidation occurs. For higher doses, the aspect remains the same [fr

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)

Thermal neutron moderating device

International Nuclear Information System (INIS)

Takigami, Hiroyuki.

1995-01-01

In a thermal neutron moderating device, superconductive coils for generating magnetic fields capable of applying magnetic fields vertical to the longitudinal direction of a thermal neutron passing tube, and superconductive coils for magnetic field gradient for causing magnetic field gradient in the longitudinal direction of the thermal neutron passing tube are disposed being stacked at the outside of the thermal neutron passing tube. When magnetic field gradient is present vertically to the direction of a magnetic moment, thermal neutrons undergo forces in the direction of the magnetic field gradient in proportion to the magnetic moment. Then, the magnetic moment of the thermal neutrons is aligned with the direction vertical to the passing direction of the thermal neutrons, to cause the magnetic field gradient in the passing direction of the thermal neutrons. The speed of the thermal neutrons can be optionally selected and the wavelength can freely be changed by applying forces to the thermal neutrons and changing the extent and direction of the magnetic field gradient. Superconductive coils are used as the coils for generating magnetic fields and the magnetic field gradient in order to change extremely high energy of the thermal neutrons. (N.H.)

Ab initio and Molecular Dynamic models of displacement damage in crystalline and turbostratic graphite

Science.gov (United States)

McKenna, Alice

One of the functions of graphite is as a moderator in several nuclear reactor designs, including the Advanced Gas-cooled Reactor (AGR). In the reactor graphite is used to thermalise the neutrons produced in the fission reaction thus allowing a self-sustained reaction to occur. The graphite blocks, acting as the moderator, are constantly irradiated and consequently suffer damage. This thesis examines the types of damage caused using molecular dynamic (MD) simulations and ab intio calculations. Neutron damage starts with a primary knock-on atom (PKA), which is travelling so fast that it creates damage through electronic and thermal excitation (this is addressed with thermal spike simulations). When the PKA has lost energy the subsequent cascade is based on ballistic atomic displacement. These two types of simulations were performed on single crystal graphite and other carbon structures such as diamond and amorphous carbon as a comparison. The thermal spike in single crystal graphite produced results which varied from no defects to a small number of permanent defects in the structure. It is only at the high energy range that more damage is seen but these energies are less likely to occur in the nuclear reactor. The thermal spike does not create damage but it is possible that it can heal damaged sections of the graphite, which can be demonstrated with the motion of the defects when a thermal spike is applied. The cascade simulations create more damage than the thermal spike even though less energy is applied to the system. A new damage function is found with a threshold region that varies with the square root of energy in excess of the energy threshold. This is further broken down in to contributions from primary and subsequent knock-on atoms. The threshold displacement energy (TDE) is found to be Ed=25eV at 300K. In both these types of simulation graphite acts very differently to the other carbon structures. There are two types of polycrystalline graphite structures

Under Water Thermal Cutting of the Moderator Vessel and Thermal Shield

International Nuclear Information System (INIS)

Loeb, A.; Sokcic-Kostic, M.; Eisenmann, B.; Prechtl, E.

2007-01-01

This paper presents the segmentation of the in 8 meter depth of water and for cutting through super alloyed moderator vessel and of the thermal shield of the MZFR stainless steel up to 130 mm wall thickness. Depending on the research reactor by means of under water plasma and contact arc metal cutting. The moderator vessel and the thermal shield are the most essential parts of the MZFR reactor vessel internals. These components have been segmented in 2005 by means of remotely controlled under water cutting utilizing a special manipulator system, a plasma torch and CAMC (Contact Arc Metal Cutting) as cutting tools. The engineered equipment used is a highly advanced design developed in a two years R and D program. It was qualified to cut through steel walls of more than 100 mm thickness in 8 meters water depth. Both the moderator vessel and the thermal shield had to be cut into such size that the segments could afterwards be packed into shielded waste containers each with a volume of roughly 1 m 3 . Segmentation of the moderator vessel and of the thermal shield was performed within 15 months. (author)

Development and prototype testing of MgCl 2 /graphite foam latent heat thermal energy storage system

Energy Technology Data Exchange (ETDEWEB)

Singh, Dileep; Yu, Wenhua; Zhao, Weihuan; Kim, Taeil; France, David M.; Smith, Roger K.

2018-01-01

Composites of graphite foam infiltrated with a magnesium chloride phase-change material have been developed as high-temperature thermal energy storage media for concentrated solar power applications. This storage medium provides a high thermal energy storage density, a narrow operating temperature range, and excellent heat transfer characteristics. In this study, experimental investigations were conducted on laboratory-scale prototypes with magnesium chloride/graphite foam composite as the latent heat thermal energy storage system. Prototypes were designed and built to monitor the melt front movement during the charging/discharging tests. A test loop was built to ensure the charging/discharging of the prototypes at temperatures > 700 degrees C. Repeated thermal cycling experiments were carried out on the fabricated prototypes, and the experimental temperature profiles were compared to the predicted results from numerical simulations using COMSOL Multiphysics software. Experimental results were found to be in good agreement with the simulations to validate the thermal models.

Time-resolved investigations of the non-thermal ablation process of graphite induced by femtosecond laser pulses

Energy Technology Data Exchange (ETDEWEB)

Kalupka, C., E-mail: christian.kalupka@llt.rwth-aachen.de; Finger, J. [Chair for Laser Technology LLT, RWTH Aachen University, Aachen 52074 (Germany); Reininghaus, M. [Chair for Laser Technology LLT, RWTH Aachen University, Aachen 52074 (Germany); Fraunhofer Institute for Laser Technology ILT, Steinbachstraße 15, Aachen 52074 (Germany)

2016-04-21

We report on the in-situ analysis of the ablation dynamics of the, so-called, laser induced non-thermal ablation process of graphite. A highly oriented pyrolytic graphite is excited by femtosecond laser pulses with fluences below the classic thermal ablation threshold. The ablation dynamics are investigated by axial pump-probe reflection measurements, transversal pump-probe shadowgraphy, and time-resolved transversal emission photography. The combination of the applied analysis methods allows for a continuous and detailed time-resolved observation of the non-thermal ablation dynamics from several picoseconds up to 180 ns. Formation of large, μm-sized particles takes place within the first 3.5 ns after irradiation. The following propagation of ablation products and the shock wave front are tracked by transversal shadowgraphy up to 16 ns. The comparison of ablation dynamics of different fluences by emission photography reveals thermal ablation products even for non-thermal fluences.

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

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)

Thermal conductivity enhancement of sodium acetate trihydrate by adding graphite powder and the effect on stability of supercooling

DEFF Research Database (Denmark)

Johansen, Jakob Berg; Dannemand, Mark; Kong, Weiqiang

2015-01-01

. The graphite powder was stabilized using carboxymetyl cellulose and successfully tested in heating and supercooling cycles with no loss of performance. Thermal conductivity enhancing properties of graphite powder was shown in samples. Since the experiments were conducted in small scale, at 200 g per sample......, large scale experiments are required to validate graphite as a thermo conductivity enhancing agent, suitable for use in seasonal heat storage applications utilizing SAT....

The Effect of Novolac and Graphite Polycrystal on the Acetone Penetration and Thermal Resistance of Nanocomposites Based on Nitrile Rubber

Directory of Open Access Journals (Sweden)

Rasool Mahboudi

2015-03-01

Full Text Available Developments of high diffusive environments in coincidence with emerging fluids with strong ability to destroy polymeric systems have resulted in rapid deformation and destruction of polymeric parts when in contact with such aggressive environments. Therefore, nowadays, there is a great need to develop highly resistant materials towards aggressive chemicals and harsh conditions. In this paper the effect of graphite polycrystal powders and novolac type phenolic resin has been experimentally studied towards acetone diffusion and thermal stability of polyacrylonitrile butadiene rubber/novolac/graphite polycrystal nanocomposites. The results obtained from dynamic mechanical thermal analysis (DMTA and swelling in acetone showed that after 32 h samples reached to 94.2% of final swelling state. By using Avrami equation and swelling experimental data, the functionality of Ln(m/m0 to novolac and graphite polycrystal weight fraction and test duration time were evaluated. This theoretical equation evaluated and predicted the amount of Ln(m/m0 with 5.92% error after 32 h. Increases in graphite polycrystal content were followed by decreases in diffusion of acetone and modulus, before glass transition temperature, and increased thermal stability and thermal resistance of the nanocomposites. Increases in novolac content by 35 wt%, decreased glass transition temperature, thermal stability and thermal resistance of the nanocomposites. In nanocomposite, containing 45 wt% of novolac, dynamic mechanical thermal analysis (DMTA data and scanning electron microscope (SEM images showed phase separation of thermoset and elastomer in the nanocomposite blend.

Decontamination of nuclear graphite by thermal processing; Dekontamination von Nukleargraphit durch thermische Behandlung

Energy Technology Data Exchange (ETDEWEB)

Florjan, Monika W.

2010-04-15

The main problem in view of the direct disposal of the nuclear graphite is its large volume. This waste contains long-lived and short-lived radionuclides which determine the waste strategy. The irradiated graphite possess high amount of the {sup 14}C isotope. The main object of the present work was the selective separation of {sup 14}C isotope from the isotope {sup 12}C by thermal treatment (pyrolysis, partial oxidation). A successful separation could reduce the radiotoxicity and offer a different disposal strategy. Three different graphite types were investigated. The samples originate from the reflector and from the flaking of spherical fuel elements of the high-temperature reactor (AVR) Juelich. The samples from the thermal column of the research reactor (Merlin, Juelich) were also investigated. The maximum tritium releases were obtained both in inert gas atmosphere (N{sub 2}) and under water vapour-oxidizing conditions at 1280 C and 900 C. Furthermore it could be shown that 28% of {sup 14}C could be released under inert gas conditions at a 1280 C. By additive of oxidizing agent such as water vapour and oxygen the {sup 14}C release could be increased. Under water vapour-oxidizing conditions at a temperature of 1280 C up to 93% of the {sup 14}C was separated from the graphite. The matrix corrosion of 5.4% was obtained. The selective separation of the {sup 14}C is possible, because a substantial part of the radiocarbon is bound near the grain boundary surfaces. (orig.)

Silver Nanoparticles and Graphitic Carbon Through Thermal Decomposition of a Silver/Acetylenedicarboxylic Salt

Directory of Open Access Journals (Sweden)

Komninou Philomela

2009-01-01

Full Text Available Abstract Spherically shaped silver nanoparticles embedded in a carbon matrix were synthesized by thermal decomposition of a Ag(I/acetylenedicarboxylic acid salt. The silver nanoparticles, which are formed either by pyrolysis at 300 °C in an autoclave or thermolysis in xylene suspension at reflux temperature, are acting catalytically for the formation of graphite layers. Both reactions proceed through in situ reduction of the silver cations and polymerization of the central acetylene triple bonds and the exact temperature of the reaction can be monitored through DTA analysis. Interestingly, the thermal decomposition of this silver salt in xylene partly leads to a minor fraction of quasicrystalline silver, as established by HR-TEM analysis. The graphitic layers covering the silver nanoparticles are clearly seen in HR-TEM images and, furthermore, established by the presence of sp2carbon at the Raman spectrum of both samples.

Interfacial Effects on the Thermal and Mechanical Properties of Graphite/Copper Composites. Final Contractor Report Ph.D. Thesis

Science.gov (United States)

Devincent, Sandra Marie

1995-01-01

Graphite surfaces are not wet by pure copper. This lack of wetting has been responsible for a debonding phenomenon that has been found in continuous graphite fiber reinforced copper matrix composites subjected to elevated temperatures. By suitably alloying copper, its ability to wet graphite surfaces can be enhanced. Information obtained during sessile drop testing has led to the development of a copper-chromium alloy that suitably wets graphite. Unidirectionally reinforced graphite/copper composites have been fabricated using a pressure infiltration casting procedure. P100 pitch-based fibers have been used to reinforce copper and copper-chromium alloys. X-ray radiography and optical microscopy have been used to assess the fiber distribution in the cast composites. Scanning electron microscopy and Auger electron spectroscopy analyses were conducted to study the distribution and continuity of the chromium carbide reaction phase that forms at the fiber/matrix interface in the alloyed matrix composites. The effects of the chromium in the copper matrix on the mechanical and thermal properties of P100Gr/Cu composites have been evaluated through tensile testing, three-point bend testing, thermal cycling and thermal conductivity calculations. The addition of chromium has resulted in an increased shear modulus and essentially zero thermal expansion in the P100Gr/Cu-xCr composites through enhanced fiber/matrix bonding. The composites have longitudinal tensile strengths in excess of 700 MPa with elastic moduli of 393 GPa. After 100 hr at 760 deg C 84 percent of the as-cast strength is retained in the alloyed matrix composites. The elastic moduli are unchanged by the thermal exposure. It has been found that problems with spreading of the fiber tows strongly affect the long transverse tensile properties and the short transverse thermal conductivity of the P100Gr/Cu-xCr composites. The long transverse tensile strength is limited by rows of touching fibers which are paths of

Numerical Investigation of the Thermal Conductivity of Graphite Nanofibers

Science.gov (United States)

Hakak Khadem, Masoud

The thermal conductivity of graphite nano-fibers (GNFs) with different styles is predicted computationally. GNFs are formed as basal planes of graphene stacked based on the catalytic configuration. The large GNF thermal conductivity relative to a base phase change material (PCM) may lead to improved PCM performance when embedded with GNFs. Three different types of GNFs are modeled: platelet, ribbon, and herringbone. Molecular dynamics (MD) simulations are used in this study as a means to predict the thermal conductivity tensor based on atomic behavior. The in-house MD code, Molecular Dynamics in Arbitrary Geometries (MDAG), was updated with the features required to create the predictions. To model both interlayer van-der Waals and intralayer covalent bonding of carbon atoms in GNFs, a combination of the optimized Tersoff potential function for atoms within the layers and a pairwise Lennard-Jones (LJ) potential function to model the interactions between the layers was used. Tests of energy conservation in the NVE ensemble have been performed to validate the employed potential model. Nose-Hoover, Andersen, and Berendsen thermostats were also incorporated into MDAG to enable MD simulations in NVT ensembles, where the volume, number of atoms, and temperature of the system are conserved. Equilibrium MD with Green-Kubo (GK) relations was then employed to extract the thermal conductivity tensor for symmetric GNFs (platelet and ribbon). The thermal conductivity of solid argon at different temperatures was calculated and compared to other studies to validate the GK implementation. Different heat current formulations, as a result of using the three-body Tersoff potential, were considered and the discrepancy in the calculated thermal conductivity values of graphene using each formula was resolved by employing a novel comparative technique that identifies the most accurate formulation. The effect of stacking configuration on the thermal conductivity of platelet and ribbon GNFs

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.

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

Aligned Carbon Nanotube Arrays Bonded to Solid Graphite Substrates: Thermal Analysis for Future Device Cooling Applications

Directory of Open Access Journals (Sweden)

Betty T. Quinton

2018-05-01

Full Text Available Carbon nanotubes (CNTs are known for high thermal conductivity and have potential use as nano-radiators or heat exchangers. This paper focuses on the thermal performance of carpet-like arrays of vertically aligned CNTs on solid graphite substrates with the idea of investigating their behavior as a function of carpet dimensions and predicting their performance as thermal interface material (TIM for electronic device cooling. Vertically aligned CNTs were grown on highly oriented pyrolytic graphite (HOPG substrate, which creates a robust and durable all-carbon hierarchical structure. The multi-layer thermal analysis approach using Netzsch laser flash analysis system was used to evaluate their performance as a function of carpet height, from which their thermal properties can be determined. It was seen that the thermal resistance of the CNT array varies linearly with CNT carpet height, providing a unique way of decoupling the properties of the CNT carpet from its interface. This data was used to estimate the thermal conductivity of individual multi-walled nanotube strands in this carpet, which was about 35 W/m-K. The influence of CNT carpet parameters (aerial density, diameter, and length on thermal resistance of the CNT carpet and its potential advantages and limitations as an integrated TIM are discussed.

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

Anisotropy in thermal conductivity of graphite flakes–SiC_p/matrix composites: Implications in heat sinking design for thermal management applications

International Nuclear Information System (INIS)

Molina, J.M.; Louis, E.

2015-01-01

Within the frame of heat dissipation for electronics, a very interesting family of anisotropic composite materials, fabricated by liquid infiltration of a matrix into preforms of oriented graphite flakes and SiC particles, has been recently proposed. Aiming to investigate the implications of the inherent anisotropy of these composites on their thermal conductivity, and hence on their potential applications, materials with matrices of Al–12 wt.% Si alloy and epoxy polymer have been fabricated. Samples have been cut at a variable angle with respect to the flakes plane and thermal conductivity has been measured by means of two standard techniques, namely, steady state technique and laser flash method. Experimental results are presented and discussed in terms of current models, from which important technological implications for heat sinking design can be derived. - Highlights: • Anisotropy in thermal conductivity of graphite flakes-based composites is evaluated. • Samples are cut in a direction forming a variable angle with the oriented flakes. • For angles 0° and 90°, thermal conductivity does not depend on sample geometry. • For intermediate angles, thermal conductivity strongly depends on sample geometry. • “Thin” samples must be thicker than 600 μm, “thick” samples must be encapsulated.

Inert annealing of irradiated graphite by inductive heating

International Nuclear Information System (INIS)

Botzem, W.; Woerner, J.

2001-01-01

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

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.

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.

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

Monte Carlo calculations on transmutation of trans-uranic nuclear waste isotopes using spallation neutrons difference of lead and graphite moderators

CERN Document Server

Hashemi-Nezhad, S R; Brandt, R; Krivopustov, M I; Kulakov, B A; Odoj, R; Sosnin, A N; Wan, J S; Westmeier, W

2002-01-01

Transmutation rates of sup 2 sup 3 sup 9 Pu and some minor actinides ( sup 2 sup 3 sup 7 Np, sup 2 sup 4 sup 1 Am, sup 2 sup 4 sup 5 Cm and sup 2 sup 4 sup 6 Cm), in two accelerator-driven systems (ADS) with lead or graphite moderating environments, were calculated using the LAHET code system. The ADS that were used had a large volume (approx 32 m sup 3) and contained no fissile material, except for a small amount of fissionable waste nuclei that existed in some cases. Calculations were performed at an incident proton energy of 1.5 GeV and the spallation target was lead. Also breeding rates of sup 2 sup 3 sup 9 Pu and sup 2 sup 3 sup 3 U as well as the transmutation rates of two long-lived fission products sup 9 sup 9 Tc and sup 1 sup 2 sup 9 I were calculated at different locations in the moderator. It is shown that an ADS with graphite moderator is a much more effective transmuter than that with lead moderator.

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

International Nuclear Information System (INIS)

Mukherjee, S.; Grigull, P.

1989-01-01

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

Thermal conductivity of an organic phase change material/expanded graphite composite across the phase change temperature range and a novel thermal conductivity model

International Nuclear Information System (INIS)

Ling, Ziye; Chen, Jiajie; Xu, Tao; Fang, Xiaoming; Gao, Xuenong; Zhang, Zhengguo

2015-01-01

Highlights: • Expanded graphite can improve thermal conductivity of RT44HC by 20–60 times. • Thermal conductivity of PCM/EG composites keeps constant before/after melting. • Thermal conductivity of PCMs nearly doubled during phase changing. • Thermal conductivity of composite PCM increases with density and percentage of EG. • The simple model predicts thermal conductivity of EG-based composites accurately. - Abstract: This work studies factors that affect the thermal conductivity of an organic phase change material (PCM), RT44HC/expanded graphite (EG) composite, which include: EG mass fraction, composite PCM density and temperature. The increase of EG mass fraction and bulk density will both enhance thermal conductivity of composite PCMs, by up to 60 times. Thermal conductivity of RT44HC/EG composites remains independent on temperature outside the phase change range (40–45 °C), but nearly doubles during the phase change. The narrow temperature change during the phase change allows the maximum heat flux or minimum temperature for heat source if attaching PCMs to a first (constant temperature) or second (constant heat flux) thermal boundary. At last, a simple thermal conductivity model for EG-based composites is put forward, based on only two parameters: mass fraction of EG and bulk density of the composite. This model is validated with experiment data presented in this paper and in literature, showing this model has general applicability to any composite of EG and poor thermal conductive materials

Thermal characterization of nitrates and nitrates/expanded graphite mixture phase change materials for solar energy storage

International Nuclear Information System (INIS)

Xiao, X.; Zhang, P.; Li, M.

2013-01-01

Highlights: • The addition of expanded graphite improved apparent thermal conductivity significantly. • The quadratic parallel model was used to predict the effective thermal conductivity. • The melting/freezing temperatures of mixture PCMs shifted slightly with adding of EG. - Abstract: Solar energy storage has become more attractive in recent years. In particular, latent thermal energy storage (LTES) with large energy storage density and isothermal heat storage/retrieval characteristics is a hot research topic. In the present study, sodium nitrate, potassium nitrate and their mixture were used as the base materials, and expanded graphite (EG) with high thermal conductivity and thermo-chemical stability was used as an additive to enhance the thermal conductivity. EG with various mass fractions was added to the base materials to form mixture phase change materials (PCMs), and the thermal characteristics of the mixtures were studied extensively. The transient hot-wire tests showed that the addition of EG enhanced the apparent thermal conductivity significantly, e.g. the apparent thermal conductivity of the nitrates/10 wt.% EG mixture PCM was increased by about 30–40%. The test results showed good agreement with theoretical calculations of the quadratic parallel model. Tests with differential scanning calorimeter (DSC) revealed that the melting/freezing temperatures of the mixture PCMs shifted slightly, compared with those of pure nitrates

Magnetic susceptibilities and thermal expansion of artificial graphites; Susceptibilites magnetiques et dilatation thermique des graphites artificiels

Energy Technology Data Exchange (ETDEWEB)

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

1960-07-01

Starting from measurements of the magnetic susceptibility made in the two principal directions of a graphite bar, the distribution function of the normals to the carbon planes in the crystallites has been evaluated. The effect of different variation in the manufacturing process on this crystalline anisotropy has been studied. From this crystalline anisotropy we have calculated the thermal expansion coefficient possessed by a compact mass of crystallites having exactly the same orientational anisotropy as the porous body consideration. The difference between this and the observed expansion coefficient leads to the determination of the expansion of the non-graphitic part of the mass which turns out to have a negative value and is also anisotropic. We have attempted to draw some conclusions from this result. (author) [French] En partant des mesures de la susceptibilite magnetique faites dans les directions des axes principaux d'un barreau de graphite, on a calcule la fonction de distribution des perpendiculaires aux plans graphitiques dans les cristallites. On a etudie les effets que pouvaient provoquer des modifications dans le procede de fabrication sur l'anisotropie cristalline. En considerant cette anisotropie cristalline, nous avons calcule le coefficient de dilatation thermique pour un bloc compact de cristallites ayant exactement la meme anisotropie d'orientation que le corps poreux en question. La difference entre cette valeur et celle mesuree du coefficient de dilatation, nous permet de calculer la dilatation pour la partie non-graphitique du bloc, en l'occurence, on trouve une valeur negative du coefficient pour cette partie, qui est egalement anisotropique. On a essaye d'en tirer quelques conclusions. (auteur)

Anisotropy in thermal conductivity of graphite flakes–SiC{sub p}/matrix composites: Implications in heat sinking design for thermal management applications

Energy Technology Data Exchange (ETDEWEB)

Molina, J.M., E-mail: jmmj@ua.es [Instituto Universitario de Materiales de Alicante, Universidad de Alicante, Ap. 99, E-03080 Alicante (Spain); Departamento de Física Aplicada, Universidad de Alicante, Ap. 99, E-03080 Alicante (Spain); Departamento de Química Inorgánica, Universidad de Alicante, Ap. 99, | E-03080 Alicante (Spain); Louis, E. [Instituto Universitario de Materiales de Alicante, Universidad de Alicante, Ap. 99, E-03080 Alicante (Spain); Departamento de Física Aplicada, Universidad de Alicante, Ap. 99, E-03080 Alicante (Spain); Unidad Asociada del Consejo Superior de Investigaciones Científicas, Universidad de Alicante, Ap. 99, | E-03080 Alicante (Spain)

2015-11-15

Within the frame of heat dissipation for electronics, a very interesting family of anisotropic composite materials, fabricated by liquid infiltration of a matrix into preforms of oriented graphite flakes and SiC particles, has been recently proposed. Aiming to investigate the implications of the inherent anisotropy of these composites on their thermal conductivity, and hence on their potential applications, materials with matrices of Al–12 wt.% Si alloy and epoxy polymer have been fabricated. Samples have been cut at a variable angle with respect to the flakes plane and thermal conductivity has been measured by means of two standard techniques, namely, steady state technique and laser flash method. Experimental results are presented and discussed in terms of current models, from which important technological implications for heat sinking design can be derived. - Highlights: • Anisotropy in thermal conductivity of graphite flakes-based composites is evaluated. • Samples are cut in a direction forming a variable angle with the oriented flakes. • For angles 0° and 90°, thermal conductivity does not depend on sample geometry. • For intermediate angles, thermal conductivity strongly depends on sample geometry. • “Thin” samples must be thicker than 600 μm, “thick” samples must be encapsulated.

Development of an apparatus for measuring the thermal conductivity of irradiated or non-irradiated graphite; Realisation d'un appareil de mesure de la conductibilite thermique du graphite irradie ou non irradie

Energy Technology Data Exchange (ETDEWEB)

Bocquet, M; Micaud, G

1962-07-01

An apparatus was developed for measuring the thermal conductivity coefficient K of irradiated or non-irradiated graphite. The measurement of K at around room temperature with an accuracy of about 6% is possible. The study specimen is placed in a vacuum between a hot and a cold source which create a temperature gradient {delta}{theta}/ {delta}x in the steady state. The amount of heat transferred, Q, is deduced from the electrical power dissipated at the hot source, after allowing for heat losses. The thermal conductivity coefficient is defined as: K = Q/S. {delta}x/{delta}{theta}, S being the cross section of the sample. Systematic studies have made it possible to determine the mean values of the thermal conductivity. (authors) [French] Un appareil de mesure du coefficient de conductibilite thermique K du graphite irradie ou non irradie a ete realise. Utilisant le principe du transfert de chaleur, il permet de mesurer K au voisinage de la temperature ambiante avec une precision de 6 pour cent environ. L'echantillon de graphite etudie est place sous vide entre une source chaude et une source froide qui creent en regime permanent un gradient de temperature {delta}{theta}/{delta}x La quantite de chaleur transferee Q est deduite de la puissance electrique dissipee dans la source chaude en deduisant les pertes thermiques. Le coefficient de conductibilite thermique est defini par: K = Q/S. {delta}x/{delta}{theta} S designant la section de l'echantillon. Des etudes systematiques ont permis de determiner pour differents graphites non irradies les valeurs moyennes des coefficients de conductibilite thermique. Ces etudes ont mis en evidence pour un type de graphite donne, l'influence de la densite apparente sur le coefficient de conductibilite thermique. A partir de mesures effectuees sur des echantillons de graphite irradies preleves par carottage dans les empilements des reacteurs a moderateur de graphite les variations du rapport K0/Ki en fonction de la dose et de la

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

International Nuclear Information System (INIS)

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

1976-01-01

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

Sintering behavior and thermal conductivity of nickel-coated graphite flake/copper composites fabricated by spark plasma sintering

Science.gov (United States)

Xu, Hui; Chen, Jian-hao; Ren, Shu-bin; He, Xin-bo; Qu, Xuan-hui

2018-04-01

Nickel-coated graphite flakes/copper (GN/Cu) composites were fabricated by spark plasma sintering with the surface of graphite flakes (GFs) being modified by Ni-P electroless plating. The effects of the phase transition of the amorphous Ni-P plating and of Ni diffusion into the Cu matrix on the densification behavior, interfacial microstructure, and thermal conductivity (TC) of the GN/Cu composites were systematically investigated. The introduction of Ni-P electroless plating efficiently reduced the densification temperature of uncoated GF/Cu composites from 850 to 650°C and slightly increased the TC of the X-Y basal plane of the GF/Cu composites with 20vol%-30vol% graphite flakes. However, when the graphite flake content was greater than 30vol%, the TC of the GF/Cu composites decreased with the introduction of Ni-P plating as a result of the combined effect of the improved heat-transfer interface with the transition layer, P generated at the interface, and the diffusion of Ni into the matrix. Given the effect of the Ni content on the TC of the Cu matrix and on the interface thermal resistance, a modified effective medium approximation model was used to predict the TC of the prepared GF/Cu composites.

Diffusion of graphite. The effect of cylindrical canals

International Nuclear Information System (INIS)

Carle, R.; Clouet d'Orval, C.; Martelly, J.; Mazancourt, T. de; Sagot, M.; Lattes, R.; Teste du Bailler, A.

1957-01-01

Experiments on thermal neutron diffusion in the graphite used as moderator in the pile G1 have been carried out. The object of these experiments is to determine: - the intrinsic quality of this graphite, characterised by its diffusion length L or its Laplacian 1/L 2 - the effect of the canals, which modifies anisotropically the macroscopic diffusion equation and is characterized by two principal diffusion regions (or two principal Laplacian), valid respectively for the diffusion in the direction of the canals and in a perpendicular direction. In order to determine them two experiments are necessary, in which the second derivatives of the flux in relation to the space coordinates are very different. These experiments form the object of the first two parts. Part 1: Diffusion along the axis of a flux coming from the pile source, and limited radially by a quasi cylindrical screen of cadmium bars. This screen, or Faraday cage is designed to give to the thermal flux produced the same radius of extrapolation to zero as that of the pile source. The determination of L (with the graphite full) has been made under the same conditions. The measurements have been interpreted in two ways. The influence of the brackets holding the detectors is discussed. Part 2: Radial diffusion in the graphite surrounding the 'long' cylindrical pile. This is well described by a sum of Bessel functions. Part 3: Results (valid for d = 1.61 t = 17 deg. C). For the graphite without cavity L = 52.7 ± 0.4 cm. The effect of the canals on the diffusion area and its anisotropy are in excellent agreement with the theory of Behrens: L(parallel) = 64.6 cm and L(perpendicular) 62.2 cm. Appendix: Theory of the Faraday cage. (author) [fr

Neutron moderation theory with thermal motion of the moderator nuclei

Energy Technology Data Exchange (ETDEWEB)

Rusov, V.D.; Tarasov, V.A.; Chernezhenko, S.A.; Kakaev, A.A.; Smolyar, V.P. [Odessa National Polytechnic University, Department of Theoretical and Experimental Nuclear Physics, Odessa (Ukraine)

2017-09-15

In this paper we present the analytical expression for the neutron scattering law for an isotropic source of neutrons, obtained within the framework of the gas model with the temperature of the moderating medium as a parameter. The obtained scattering law is based on the solution of the general kinematic problem of elastic scattering of neutrons on nuclei in the L-system. Both the neutron and the nucleus possess arbitrary velocities in the L-system. For the new scattering law we obtain the flux densities and neutron moderation spectra as functions of temperature for the reactor fissile medium. The expressions for the moderating neutrons spectra allow reinterpreting the physical nature of the underlying processes in the thermal region. (orig.)

Pyrolysis and its potential use in nuclear graphite disposal

International Nuclear Information System (INIS)

Mason, J.B.; Bradbury, D.

2001-01-01

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

Exploration of coal-based pitch precursors for ultra-high thermal conductivity graphite fibers. Final report

Energy Technology Data Exchange (ETDEWEB)

Deshpande, G.V. [Amoco Performance Products, Inc., Alpharetta, GA (United States)

1996-12-27

Goal was to explore the utility of coal-based pitch precursors for use in ultra high thermal conductivity carbon (graphite) fibers. From graphite electrode experience, it was established that coal-based pitches tend to form more highly crystalline graphite at lower temperatures. Since the funding was limited to year 1 effort of the 3 year program, the goal was only partially achieved. The coal-base pitches can form large domain mesophase in spite of high N and O contents. The mesophase reactivity test performed on one of the variants of coal-based pitch (DO84) showed that it was not a good candidate for carbon fiber processing. Optimization of WVU`s isotropic pitch process is required to tailor the pitch for carbon fiber processing. The hetero atoms in the coal pitch need to be reduced to improve mesophase formation.

An automatic regulating control system for a graphite moderated reactor using digital techniques

International Nuclear Information System (INIS)

Carvalho Goncalves Junior, J. de.

1989-01-01

The work propose an automatic regulating control system for a graphite moderated reactor using digital techniques. The system uses a microcomputer to monitor the power and the period, to run the control algorithm, and to generate electronic signals to excite the motor, which moves vertically the control rod banks. A nuclear reactor simulator was developed to test the control system. The simulator consists of a software based on the point kinetic equations and implanted in an analogical computer. The results show that this control system has a good performance and versatility. In addition, the simulator is capable of reproducing with accuracy the behavior of a nuclear reactor. (author)

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.

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

A Study of the Oxidation Behaviour of Pile Grade A (PGA) Nuclear Graphite Using Thermogravimetric Analysis (TGA), Scanning Electron Microscopy (SEM) and X-Ray Tomography (XRT).

Science.gov (United States)

Payne, Liam; Heard, Peter J; Scott, Thomas B

2015-01-01

Pile grade A (PGA) graphite was used as a material for moderating and reflecting neutrons in the UK's first generation Magnox nuclear power reactors. As all but one of these reactors are now shut down there is a need to understand the residual state of the material prior to decommissioning of the cores, in particular the location and concentration of key radio-contaminants such as 14C. The oxidation behaviour of unirradiated PGA graphite was studied, in the temperature range 600-1050°C, in air and nitrogen using thermogravimetric analysis, scanning electron microscopy and X-ray tomography to investigate the possibility of using thermal degradation techniques to examine 14C distribution within irradiated material. The thermal decomposition of PGA graphite was observed to follow the three oxidation regimes historically identified by previous workers with limited, uniform oxidation at temperatures below 600°C and substantial, external oxidation at higher temperatures. This work demonstrates that the different oxidation regimes of PGA graphite could be developed into a methodology to characterise the distribution and concentration of 14C in irradiated graphite by thermal treatment.

Effects of core models and neutron energy group structures on xenon oscillation in large graphite-moderated reactors

International Nuclear Information System (INIS)

Yamasita, Kiyonobu; Harada, Hiroo; Murata, Isao; Shindo, Ryuichi; Tsuruoka, Takuya.

1993-01-01

Xenon oscillations of large graphite-moderated reactors have been analyzed by a multi-group diffusion code with two- and three-dimensional core models to study the effects of the geometric core models and the neutron energy group structures on the evaluation of the Xe oscillation behavior. The study clarified the following. It is important for accurate Xe oscillation simulations to use the neutron energy group structure that describes well the large change in the absorption cross section of Xe in the thermal energy range of 0.1∼0.65 eV, because the energy structure in this energy range has significant influences on the amplitude and the period of oscillations in power distributions. Two-dimensional R-Z models can be used instead of three-dimensional R-θ-Z models for evaluation of the threshold power of Xe oscillation, but two-dimensional R-θ models cannot be used for evaluation of the threshold power. Although the threshold power evaluated with the R-θ-Z models coincides with that of the R-Z models, it does not coincide with that of the R-θ models. (author)

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.

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

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

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

Final report on graphite irradiation test OG-2

International Nuclear Information System (INIS)

Price, R.J.; Beavan, L.A.

1975-01-01

Results are presented of dimensional, thermal expansivity, thermal conductivity, Young's modulus, and tensile strength measurements on specimens of nuclear graphites irradiated in capsule OG-2. About half the irradiation space was allocated to H-451 near-isotropic petroleum-coke-based graphite or its subsized prototype grade H-429. Most of these specimens had been previously irradiated. Virgin specimens of another near-isotropic graphite, grade TS-1240, were irradiated. Some previously irradiated specimens of needle-coke-based H-327 graphite and pitch-coke-based P 3 JHAN were also included

Technique for production of graphite-carbon products

Energy Technology Data Exchange (ETDEWEB)

Antonov, A.N.; Bentsianovskaya, I.A.; Filatova, V.A.; Nabokov, V.S.; Nestor, V.P.; Zil' bergleyt, I.M.

1982-01-01

The technique for producing carbon-graphite products that includes filtration under a pressure of 0.1-015 MPa (through graphite stock) of an aqueous carbon material with the addition of surfactant, drying, and subsequent thermal treatment, is simplified and made less lengthy. Oxidized graphite is utilized with a prior addition of 1-10% water-soluble organic substance into the suspension -molasses, hemicellulose, sugar or polyacrylamide. A 0.03-1.5% suspension of oxidized graphite is utilized, with a particle size of 0.02-0.1 mkm. Thermal processing is done in a carbon fill, at a rate of 10-20 degrees/hour to 700-800/sup 0/, maintained 2-3 hours.

Minor Actinide Burn in Thermal Spectrum with Enhanced Moderation

International Nuclear Information System (INIS)

Petrovic, B.; Huang, L. M.

2010-01-01

Resolving the issue of spent nuclear fuel and nuclear waste management is the necessary condition for long-term sustainability of nuclear power, and requires addressing plutonium, minor actinides (MA) and fission products. Various strategies from once-through homogeneous burn to partitioning and transmutation, and from thermal to fast systems, are being considered. The optimum system-level performance will likely require advanced critical or subcritical systems with a range of neutron spectra. Thermal systems, while not optimum, may be deployed sooner, and may provide mid-term amelioration of the issue. This paper examines burn of MA in thermal systems. One specific concern in this case is deterioration of safety parameters due to a high thermal absorption cross section of MA. Enhanced moderation has potential to at least partly remedy this concern. Therefore, we have evaluated adopting the IRIS neutronic design to MA burn. The IRIS reactor design offers enhanced safety margin, due to its fully passive safety systems and safety-by-design approach. Also, in addition to the standard UO 2 fuel (reference IRIS design), an alternative core with enhanced moderation fuel was considered. These two features (safety margin, enhanced moderation) provide a good starting point for MA burn in a thermal system. Further modifications to accommodate MA-bearing rods will be discussed. The paper will examine the benefit of the enhanced moderation in comparison to homogeneous MA burn in a typical PWR reactor.(author).

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.

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

Thermophysical properties estimation of paraffin/graphite composite phase change material using an inverse method

International Nuclear Information System (INIS)

Lachheb, Mohamed; Karkri, Mustapha; Albouchi, Fethi; Mzali, Foued; Nasrallah, Sassi Ben

2014-01-01

Highlights: • Preparation of paraffin/graphite composites by uni-axial compression technique. • Measurement of thermophysical properties of paraffin/graphite using the periodic method. • Measurement of the experimental densities of paraffin/graphite composites. • Prediction of the effective thermal conductivity using analytical models. - Abstract: In this paper, two types of graphite were combined with paraffin in an attempt to improve thermal conductivity of paraffin phase change material (PCM): Synthetic graphite (Timrex SFG75) and graphite waste obtained from damaged Tubular graphite Heat Exchangers. These paraffin/graphite phase change material (PCM) composites are prepared by the cold uniaxial compression technique and the thermophysical properties were estimated using a periodic temperature method and an inverse technique. Results showed that the thermal conductivity and thermal diffusivity are greatly influenced by the graphite addition

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

Preparation, characterization and thermal properties of binary nitrate salts/expanded graphite as composite phase change material

Energy Technology Data Exchange (ETDEWEB)

Xiao, Junbing [School of Materials and Energy, Guangdong University of Technology, 510006 Guangzhou (China); Huang, Jin, E-mail: huangjiner@126.com [School of Materials and Energy, Guangdong University of Technology, 510006 Guangzhou (China); Zhu, Panpan; Wang, Changhong [School of Materials and Energy, Guangdong University of Technology, 510006 Guangzhou (China); Li, Xinxi [School of Materials and Energy, Guangdong University of Technology, 510006 Guangzhou (China); Center for Nanochemistry, Beijing National Laboratory for Molecular Sciences, State Key Laboratory for Structural Chemistry of Unstable and Stable Species, College of Chemistry and Molecular Engineering, Peking University, Beijing (China)

2014-07-01

Highlights: • The expanded graphite enhanced thermal conductivity coefficient greatly. • The aqueous solution method adopting ultrasonic was utilized to disperse EG. • The combination of composite was physical without chemical reaction. • The reduction on total latent heat was slight after the adding EG. - Abstract: The binary nitrate salts/expanded graphite (EG) composite phase change material (PCM) were prepared via adding different mass rate of EG to binary nitrate salts consisting of NaNO{sub 3} and KNO{sub 3} (6:4) by aqueous solution method adopting ultrasonic. The morphology and chemical composition of EG and the composite PCM were characterized and investigated by X-ray diffraction (XRD), scan electron microscope (SEM), energy dispersive spectrometer (EDS), transmission electron microscope (TEM), respectively. Laser thermal conductivity instrument and differential scanning calorimeter (DSC) were employed to measure thermo physical properties. Drawing the conclusion from investigation, that EG had enhanced thermal conductivity coefficient which largely increased to 4.884 W/(m K) and reduced total latent heat by mostly 11.0%. The morphology and phase structure results indicated that EG were well dispersed into and physically combined with molten salts. In general, the prepared composite PCM could be a suitable phase change material for thermal energy storage.

Preparation, characterization and thermal properties of binary nitrate salts/expanded graphite as composite phase change material

International Nuclear Information System (INIS)

Xiao, Junbing; Huang, Jin; Zhu, Panpan; Wang, Changhong; Li, Xinxi

2014-01-01

Highlights: • The expanded graphite enhanced thermal conductivity coefficient greatly. • The aqueous solution method adopting ultrasonic was utilized to disperse EG. • The combination of composite was physical without chemical reaction. • The reduction on total latent heat was slight after the adding EG. - Abstract: The binary nitrate salts/expanded graphite (EG) composite phase change material (PCM) were prepared via adding different mass rate of EG to binary nitrate salts consisting of NaNO 3 and KNO 3 (6:4) by aqueous solution method adopting ultrasonic. The morphology and chemical composition of EG and the composite PCM were characterized and investigated by X-ray diffraction (XRD), scan electron microscope (SEM), energy dispersive spectrometer (EDS), transmission electron microscope (TEM), respectively. Laser thermal conductivity instrument and differential scanning calorimeter (DSC) were employed to measure thermo physical properties. Drawing the conclusion from investigation, that EG had enhanced thermal conductivity coefficient which largely increased to 4.884 W/(m K) and reduced total latent heat by mostly 11.0%. The morphology and phase structure results indicated that EG were well dispersed into and physically combined with molten salts. In general, the prepared composite PCM could be a suitable phase change material for thermal energy storage

Distribution of the thermal neutron field around the graphite reflector of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Khang, Ngo Phu; Binh, Nguyen Duc; Tuan, Nguyen Minh; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

Thermal neutron flux distributions around the graphite reflector of the Dalat Nuclear Research Reactor are determined by the method for neutron activating Cu foils. The major results are as follows: a/The axial distributions at the inner and outer margins of the graphite reflector have unsymmetrical shapes, similar to axial distributions in the core. There is a dissimilarity between the distribution curves at the inner margin and those at the outer margin of the reflector. b/ The radial distribution on the upper surface of the graphite reflector is measured and is described by the two-group neutron diffusion theory. The maximal value of the curve lies at the position of R{sub m}ax = 22.5 cm. c/ The distribution in the twenty water irradiation holes around the rotary specimen rack is obtained. (author). 3 refs., 5 figs., 1 tab.

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

International Nuclear Information System (INIS)

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

1963-01-01

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

Irradiation damage in graphite. The works of Professor B.T. Kelly

International Nuclear Information System (INIS)

Marsden, B.J.

1996-01-01

The irradiation damage produced in graphite by energetic neutrons (>100eV) has been extensively studied because of the use of graphite as a moderator in thermal nuclear reactors. In recent times, graphite has been adopted as the protective tiling of the inner wall of experimental fusion systems and property changes due to fusion neutrons have become important. The late Professor B.T. Kelly reviewed the work carried out on the irradiation behaviour of graphite since the 1940s. This work is particularly timely as the scale of research into the effects of fission neutrons has been greatly reduced and many of the active researchers have retired. In recent years, new programmes of work are being formulated for the use of graphite in both the field of high temperature reactor systems and fusion systems. It is therefore important that the knowledge gained by Professor Kelly and other workers is not lost but passed on to future generations of nuclear scientists and engineers. This paper reviews Professor Kelly's last work, it also draws on the experience gained during many long discussions with Brian during the years he worked closely with the present graphite team at AEA Technology. It is hoped to publish his work in full in the near future. (author). 13 refs, 14 figs, 3 tabs

Influence of different moderator materials on characteristics of neutron fluxes generated under irradiation of lead target with proton beams

International Nuclear Information System (INIS)

Sosnin, A.N.; Polanski, A.; Petrochenkov, S.A.

2002-01-01

Neutron fields generated in extended heavy (Z ≥ 82) targets under irradiation with proton beams at energies in the range of 1 GeV are investigated. Influence of different moderators on the spectra and multiplicities of neutrons escaping the surface of the assembly consisting of a lead target (diam. 8 cm x 20 cm or diam. 8 cm x 50 cm) screened by variable thickness of polyethylene or graphite, respectively, was compared. It is shown that the effectiveness of graphite as a material used in such assemblies to moderate spallation neutrons down to thermal energies is significantly lower than that of paraffin

Thickness optimization of various moderator materials for maximization of thermal neutron fluence

International Nuclear Information System (INIS)

Dhang, Prosenjit; Verma, Rishi; Shyam, Anurag

2015-01-01

Plasma focus device is widely being used as pulsed neutron source for variety of applications. Measurements of neutron yield by largely preferred Helium-3 proportional counter and Silver activation counter are mainly sensitive to thermal neutrons and are typically used with a neutron moderator. Thermalization of neutron is based on scattering reaction and hydrogenous materials are the best thermalizing medium. The efficiency of aforementioned neutron detectors is considerably affected by physical and geometrical properties of thermalizing medium i.e. moderator material, its thickness and shape. In view of the same, simulations have been performed to explore the effective utilization of Polyethylene, Perspex and Light water as moderating mediums for cylindrical and spherical geometry. In this study, estimated thermal fluence value up to 0.5 eV has been considered as the benchmark factor for comparing efficient thermalization by specific material, its thickness and shape. In either of the shapes being cylindrical or spherical, use of Polyethylene as moderating medium has resulted in minimum optimum thickness along with highest thermal fluence. (author)

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.

Final report on graphite irradiation test OG-3

International Nuclear Information System (INIS)

Price, R.J.; Beavan, L.A.

1977-01-01

The results of dimensional, thermal expansivity, thermal conductivity, Young's modulus, and tensile strength measurements on graphite specimens irradiated in capsule OG-3 are presented. The graphite grades investigated included near-isotropic H-451 (three different preproduction lots), TS-1240, and SO818; needle coke H-327; and European coal tar pitch coke grades P 3 JHA 2 N, P 3 JHAN, and ASI2-500. Data were obtained in the temperature range 823 0 K to 1673 0 K. The peak fast neutron fluence in the experiment was 3 x 10 25 n/m 3 (E greater than 29 fJ)/sub HTGR/; the total accumulated fluence exceeded 9 x 10 25 n/m 2 on some H-451 specimens and 6 x 10 25 n/m 2 on some TS-1240 specimens. Irradiation-induced dimensional changes on H-451 graphite differed slightly from earlier predictions. For an irradiation temperature of about 1225 0 K, axial shrinkage rates at high fluences were somewhat higher than predicted, and the fluence at which radial expansion started (about 9 x 10 25 n/m 2 at 1275 0 K) was lower. TS-1240 graphite underwent smaller dimensional changes than H-451 graphite, while limited data on SO818 and ASI2-500 graphites showed similar behavior to H-451. P 3 JHAN and P 3 JHA 2 N graphites displayed anisotropic behavior with rapid axial shrinkage. Comparison of dimensional changes between specimens from three logs of H-451 and of TS-1240 graphites showed no significant log-to-log variations for H-451, and small but significant log-to-log variations for TS-1240. The thermal expansivity of the near-isotropic graphites irradiated at 865-1045 0 K first increased by 5 percent to 10 percent and then decreased. At higher irradiation temperatures the thermal expansivity decreased by up to 50 percent. Changes in thermal conductivity were consistent with previously established curves. Specimens which were successively irradiated at two different temperatures took on the saturation conductivity for the new temperature

Correlation between some mechanical and physical properties of polycrystalline graphites

International Nuclear Information System (INIS)

Yoda, Shinichi; Fujisaki, Katsuo

1982-01-01

Mechanical and physical properties of polycrystalline graphites, tensile strength, compressive strength, flexural strength, Young's modulus, thermal expansion coefficient, electrical resistivity, volume fraction of porosity, and graphitisation were measured for ten brand graphites. Correlation between the mechanical and physical properties of the graphites were studied. Young's modulus and thermal expansion coefficient of the graphites depend on volume fraction of porosity. The Young's modulus of the graphites tended to increase with increasing the thermal expansion coefficient. For an anisotropic graphite, an interesting relationship between the Young's modulus E and the thermal expansion coefficient al pha was found in any specimen orientations; alpha E=constant. The value of alphah E was dependent upon the volume fraction of porosity. It should be noted here that the electrical resistivity increased with decreasing grain size. The flexural and the compressive strength were related with the volume fraction of porosity while the tensile strength was not, The relationships between the tensile, the compressive and the flexural strength can be approximately expressed as linear functions over a wide range of the stresses. (author)

Thermal analysis of CANDU6 moderator system for loss of cooling

International Nuclear Information System (INIS)

Xu Zhen

2012-01-01

The coolant system and moderator system of CANDU6 are independent. The prompt neutrons are moderated as thermal neutrons by the moderator and the continuous nuclear fission in the reactor is maintained. At the same time the moderator system supplies the heat sink for the heat produced by the neutrons moderation. During the in-service maintenance of plant, the standby RCW which will only cool down reactor coolant system operates instead of RCW and can not supply heat sink for moderator system heat exchanger. As the result, the moderator system will lose heat sink during the operation of standby RCW. To estimate the moderator temperature, the thermal analysis of moderator system for loss of cooling was compared with the experiment data and the system failure caused by the temperature raising was evaluated in this paper. (author)

Quantifying the relevance of adaptive thermal comfort models in moderate thermal climate zones

Energy Technology Data Exchange (ETDEWEB)

Hoof, Joost van; Hensen, Jan L.M. [Faculty of Architecture, Building and Planning, Technische Universiteit Eindhoven, Vertigo 6.18, P.O. Box 513, 5600 MB Eindhoven (Netherlands)

2007-01-15

Standards governing thermal comfort evaluation are on a constant cycle of revision and public review. One of the main topics being discussed in the latest round was the introduction of an adaptive thermal comfort model, which now forms an optional part of ASHRAE Standard 55. Also on a national level, adaptive thermal comfort guidelines come into being, such as in the Netherlands. This paper discusses two implementations of the adaptive comfort model in terms of usability and energy use for moderate maritime climate zones by means of literature study, a case study comprising temperature measurements, and building performance simulation. It is concluded that for moderate climate zones the adaptive model is only applicable during summer months, and can reduce energy for naturally conditioned buildings. However, the adaptive thermal comfort model has very limited application potential for such climates. Additionally we suggest a temperature parameter with a gradual course to replace the mean monthly outdoor air temperature to avoid step changes in optimum comfort temperatures. (author)

Thermal conductivity predictions of herringbone graphite nanofibers using molecular dynamics simulations.

Science.gov (United States)

Khadem, Masoud H; Wemhoff, Aaron P

2013-02-28

Non-equilibrium molecular dynamics (NEMD) simulations are used to investigate the thermal conductivity of herringbone graphite nanofibers (GNFs) at room temperature by breaking down the axial and transverse conductivity values into intralayer and interlayer components. The optimized Tersoff potential is used to account for intralayer carbon-carbon interactions while the Lennard-Jones potential is used to model the interlayer carbon-carbon interactions. The intralayer thermal conductivity of the graphene layers near room temperature is calculated for different crease angles and number of layers using NEMD with a constant applied heat flux. The edge effect on a layer's thermal conductivity is investigated by computing the thermal conductivity values in both zigzag and armchair directions of the heat flow. The interlayer thermal conductivity is also predicted by imposing hot and cold Nosé-Hoover thermostats on two layers. The limiting case of a 90° crease angle is used to compare the results with those of single-layer graphene and few-layer graphene. The axial and transverse thermal conductivities are then calculated using standard trigonometric conversions of the calculated intralayer and interlayer thermal conductivities, along with calculations of few-layer graphene without a crease. The results show a large influence of the crease angle on the intralayer thermal conductivity, and the saturation of thermal conductivity occurs when number of layers is more than three. The axial thermal conductivity, transverse thermal conductivity in the crease direction, and transverse thermal conductivity normal to the crease for the case of a five-layer herringbone GNF with a 45° crease angle are calculated to be 27 W∕m K, 263 W∕m K, and 1500 W∕m K, respectively, where the axial thermal conductivity is in good agreement with experimental measurements.

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

Structures and Performance of Graphene/Polyimide Composite Graphite Fibers

Directory of Open Access Journals (Sweden)

LI Na

2017-09-01

Full Text Available Dry-wet spinning process was used to gain graphene oxide/polyimide composite fibers, then graphene/polyimide composite carbon and graphite fibers were obtained through carbonized and graphitized. Different graphene oxide contents of the composite carbon and graphite fibers were measured by thermal gravimetric analysis, Raman, mechanical properties, electrical properties,SEM and so on. The results show that when the GO content is 0.3%(mass fraction,the same below, the thermal property of the graphene oxide/polyimide composite fibers is the best. The mechanical and electrical properties are obriously improved by the addition of GO, graphitization degree also increases. When the composite carbon fibers are treated at 2800℃, GO content increases to 2.0%, the thermal conductivity of the composite graphite fibers reaches 435.57W·m-1·K-1 and cross-section structures of carbon fibers are more compact.

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...

Thermal-hydraulic analysis of graphite tubes for the non-aqueous system of accelerator transmutation of nuclear waste

International Nuclear Information System (INIS)

Potter, R.C.; Venneri, F.; Trujillo, D.A.

1993-01-01

Accelerator transmutation of nuclear waste offers exciting possibilities for the disposal of nuclear waste by converting it into more benign Species. The non-aqueous system discussed here contains the materials to be transmuted within a lithium-fluoride salt. The system consists of bundles of graphite tubes containing the salt Solution. The tubes are cooled as lithium flows across their exterior. These circular graphite tubes have an inner circular passage and an outer annulus. Natural convection within the tubes causes the salt to circulate. This paper deals with the thermal-hydraulics of the system; it does not consider the neutronics in detail. Heat transfer and fluid flow were modeled using a custom computer program the system behavior of an graphite tube. Different geometries were tried, while keeping the system volume the same, to determine an optimize graphite tube geometry. I considered both the parallel flow and the counterflow of the lithium coolant, and allowed limited boiling to occur to facilitate circulation. I achieved power densities as high as 200 W/cm 3 for the overall blanket

Design of hyper-thermal neutron irradiation fields for neutron capture therapy in KUR-heavy water neutron irradiation facility. Mounting of hyper-thermal neutron converter in therapeutic collimator

International Nuclear Information System (INIS)

Sakurai, Y.; Kobayashi, T.

2001-01-01

Neutron capture therapy (NCP) using thermal neutron needs to improve of depth dose distribution in a living body. Epi-thermal neutron following moderation of fast neutron is usually used for improving of the depth dose distribution. The moderation method of fast neutron, however, gets mixed some of high energy neutron which give some of serious effects to a living body, and involves the difficulty for collimation of thermal neutron to the diseased part. Hyper-thermal neutrons, which are in an energy range of 0.1-3 eV at high temperature side of thermal neutron, are under consideration for application to the NCP. The hyper-thermal neutrons can be produced by up-scattering of thermal neutron in a high temperature material. Fast neutron components in collimator for the NCP reduce on application of the up-scattering method. Graphite at high temperature (>1000k) is used as a hyper-thermal neutron converter. The hyper-thermal neutron converter is planted to mount on therapeutic collimator which is located at the nearest side of patient for the NCP. Total neutron flux, ratio of hyper-thermal neutron to total neutron, and ratio of gamma-ray dose to neutron flux are calculated as a function of thickness of the graphite converter using monte carlo code MCNP-V4B. (M. Suetake)

Thermal-mechanical properties of a graphitic-nanofibers reinforced epoxy.

Science.gov (United States)

Salehi-Khojin, Amin; Jana, Soumen; Zhong, Wei-Hong

2007-03-01

We previously developed a series of reactive graphitic nanofibers (r-GNFs) reinforced epoxy (nano-epoxy) as composite matrices, which have shown good wetting and adhesion properties with continuous fiber. In this work, the thermal-mechanical properties of the nano-epoxy system containing EponTM Resin 828 and Epi-cure Curing Agent W were characterized. Results from three-point bending tests showed that the flexural strength and flexural modulus of this system with 0.30 wt% of reactive nanofibers were increased by 16%, and 21% respectively, over pure epoxy. Fracture toughness increased by ca. 40% for specimens with 0.50 wt% of r-GNFs. By dynamic mechanical analysis (DMA) test, specimens with 0.30 wt% of r-GNFs showed a significant increase in storage modulus E' (by ca. 122%) and loss modulus E" (by ca. 111%) with respect to that of pure epoxy. Also thermo-dilatometry analysis (TDA) was used to measure dimensional change of specimens as a function of temperature, and then, coefficients of thermal expansion (CTE) before and after glass transition temperature (Tg) were obtained. Results implied that nano-epoxy materials had good dimensional stability and reduced CTE values when compared to those of pure epoxy.

Measurement of thermal neutron spectra using LINAC in Japan Atomic Energy Research Institute (JAERI)

International Nuclear Information System (INIS)

Akino, Fujiyoshi

1982-01-01

The exact grasp of thermal neutron spectra in a core region is very important for obtaining accurate thermal neutron group constants in the calculation for the nuclear design of a reactor core. For the accurate grasp of thermal neutron spectra, the capability of thermal neutron spectra to describe the moderator cross-sections for thermal neutron scattering is a key factor. Accordingly, 0 deg angular thermal neutron spectra were measured by the time of flight (TOF) method using the JAERI LINAC as a pulsed neutron source, for light water system added with Cd and In, high temperature graphite system added with boron, and light water-natural uranium heterogeneous multiplication system among the reactor moderators of light water or graphite systems. First, the equations to give the time of flight and neutron flux by TOF method were analyzed, and several corrections were investigated, such as those for detector efficiency, background, the transmission coefficient of air and the Al window of a flight tube, mean emission time of neutrons, and the distortion effect of re-entrant hole on thermal neutron spectra. Then, the experimental system, results and calculation were reported for the experiments on the above three moderator systems. Finally, the measurement of fast neutron spectra in natural uranium system and that of the efficiency of a 6 Li glass scintillator detector are described. (Wakatsuki, Y.)

Graphite-high density polyethylene laminated composites with high thermal conductivity made by filament winding

Directory of Open Access Journals (Sweden)

W. Lv

2018-03-01

Full Text Available The low thermal conductivity of polymers limits their use in numerous applications, where heat transfer is important. The two primary approaches to overcome this limitation, are to mix in other materials with high thermal conductivity, or mechanically stretch the polymers to increase their intrinsic thermal conductivity. Progress along both of these pathways has been stifled by issues associated with thermal interface resistance and manufacturing scalability respectively. Here, we report a novel polymer composite architecture that is enabled by employing typical composites manufacturing method such as filament winding with the twist that the polymer is in fiber form and the filler in form of sheets. The resulting novel architecture enables accession of the idealized effective medium composite behavior as it minimizes the interfacial resistance. The process results in neat polymer and 50 vol% graphite/polymer plates with thermal conductivity of 42 W·m–1·K–1 (similar to steel and 130 W·m–1·K–1 respectively.

Influence of Different Moderator Materials on Characteristics of Neutron Fluxes Generated under Irradiation of Lead Target with Proton Beams

CERN Document Server

Sosnin, A N; Polanski, A; Petrochenkov, S A; Golovatyuk, V M; Krivopustov, M I; Bamblevski, V P; Westmeier, W; Odoj, R; Brandt, R; Robotham, H; Hashemi-Nezhad, S R; Zamani-Valassiadou, M

2002-01-01

Neutron fields generated in extended heavy (Z\\geq 82) targets under irradiation with proton beams at energies in the range of 1 GeV are investigated. Influence of different moderators on the spectra and multiplicities of neutrons escaping the surface of the assembly consisting of a lead target (\\varnothing 8 cm\\times 20 cm or \\varnothing 8cm\\times 50 cm) screened by variable thickness of polyethylene or graphite, respectively, was compared in the present work. It is shown that the effectiveness of graphite as a material used in such assemblies to moderate spallation neutrons down to thermal energies is significantly lower than that of paraffin.

Determination of thermal neutrons diffusion length in graphite; Determinacion de la Longitud de Difusion de los Neutrones Termicos en Grafito

Energy Technology Data Exchange (ETDEWEB)

Garcia Fite, J

1959-07-01

The diffusion length of thermal neutrons in graphite using the less possible quantity of material has been determined. The proceeding used was the measurement in a graphite pile which has a punctual source of rapid neutrons inside surrounded by a reflector medium (paraffin or water). The measurement was done in the following conditions: a) introducing an aluminium plate between both materials. b) Introducing a cadmium plate between both materials. (Author) 91 refs.

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

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)

Thermal desorption spectroscopy of pyrolytic graphite cleavage faces after keV deuterium irradiation at 330-1000 K

International Nuclear Information System (INIS)

Gotoh, Y.; Yamaki, T.; Tokiguchi, K.

1992-01-01

Thermal desorption spectroscopy (TDS) measurements were made on D 2 and CD 4 from surface layers of pyrolytic graphite cleavage faces after 3 keV D + 3 irradiation to 1.5 x 10 18 D/cm 2 at irradiation temperatures from 330 to 1000 K. Thermal desorption of both D 2 and CD 4 was observed to rise simultaneously at around 700 K. The D 2 peak was found at T m = 900-1000 K, while the CD 4 peak appeared at a lower temperature, 800-840 K. The T m for the D 2 TDS increased, while that for the CD 4 decreased with increasing irradiation temperature. These results obviously indicate that the D 2 desorption is detrapping/recombination limited, while the CD 4 desorption is most likely to be diffusion limited. The amount of thermally desorbed D 2 after the D + irradiation was observed to monotonously decrease as the irradiation temperature was increased from 330 to 1000 K. These tendencies agreed with previous results for the irradiation temperature dependencies of both C1s chemical shift (XPS) and the interlayer spacing, d 002 (HRTEM), on the graphite basal face. (orig.)

Thermal Characterization of Lauric-Stearic Acid/Expanded Graphite Eutectic Mixture as Phase Change Materials.

Science.gov (United States)

Zhu, Hua; Zhang, Peng; Meng, Zhaonan; Li, Ming

2015-04-01

The eutectic mixture of lauric acid (LA) and stearic acid (SA) is a desirable phase change material (PCM) due to the constant melting temperature and large latent heat. However, its poor thermal conductivity has hampered its broad utilization. In the present study, pure LA, SA and the mixtures with various mass fractions of LA-SA were used as the basic PCMs, and 10 wt% expanded graphite (EG) was added to enhance the thermal conductivities. The phase change behaviors, microstructural analysis, thermal conductivities and thermal stabilities of the mixtures of PCMs were investigated by differential scanning calorimetry (DSC), scanning electronic microscope (SEM), transient plane source (TPS) and thermogravimetric analysis (TGA), respectively. The results show that the LA-SA binary mixture of mixture ratio of 76.3 wt%: 23.7 wt% forms an eutectic mixture, which melts at 38.99 °C and has a latent heat of 159.94 J/g. The melted fatty acids are well absorbed by the porous network of EG and they have a good thermal stability. Furthermore, poor thermal conductivities can be well enhanced by the addition of EG.

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

Energy Technology Data Exchange (ETDEWEB)

Santha Kumar, Arunjunai Raja Shankar [Materials Science Centre, Indian Institute of Technology, Kharagpur, 721302, West Bengal (India); Leibniz-Institut für Polymerforschung Dresden e.V., Hohe Straße 6, 01069, Dresden (Germany); Piana, Francesco [Leibniz-Institut für Polymerforschung Dresden e.V., Hohe Straße 6, 01069, Dresden (Germany); Organic Chemistry of Polymers, Technische Universität Dresden, 01062, Dresden (Germany); Mičušík, Matej [Polymer Institute, Slovak Academy of Sciences, Dúbravská cesta 9, 845 41, Bratislava (Slovakia); Pionteck, Jürgen, E-mail: pionteck@ipfdd.de [Leibniz-Institut für Polymerforschung Dresden e.V., Hohe Straße 6, 01069, Dresden (Germany); Banerjee, Susanta [Materials Science Centre, Indian Institute of Technology, Kharagpur, 721302, West Bengal (India); Voit, Brigitte [Leibniz-Institut für Polymerforschung Dresden e.V., Hohe Straße 6, 01069, Dresden (Germany); Organic Chemistry of Polymers, Technische Universität Dresden, 01062, Dresden (Germany)

2016-10-01

Heavily oxidized and ordered graphene nanoplatelets were produced from natural graphite by oxidation using a mixture of phosphoric acid, sulphuric acid, and potassium permanganate (Marcano's method). The atomic percentage of oxygen in the graphite oxide produced was more than 30% confirmed by XPS studies. The graphite oxide produced had intact basal planes and remains in a layered structure with interlayer distance of 0.8 nm, analyzed by WAXS. The graphite oxide was treated with 4,4′-methylenebis(phenyl isocyanate) (MDI) to produce grafted isocyanate functionalization. Introduction of these bulky functional groups widens the interlayer distance to 1.3 nm. In addition, two reduction methods, namely benzyl alcohol mediated reduction and thermal reduction were carried out on isocyanate modified and unmodified graphite oxides and compared to each other. The decrease in the oxygen content and the sp{sup 3} defect-repair were studied with XPS and RAMAN spectroscopy. Compared to the thermal reduction process, which is connected with large material loss, the benzyl alcohol mediated reduction process is highly effective in defect repair. This resulted in an increase of conductivity of at least 9 orders of magnitude compared to the graphite oxide. - Highlights: • Preparation of GO by Marcano's method results in defined interlayer spacing. • Treatment of GO with diisocyanate widens the interlayer spacing to 1.3 nm. • Chemical reduction of GO with benzyl alcohol is effective in defect repair. • Electrical conductivity increases by 9 orders of magnitude during chemical reduction. • The isocyanate functionalization is stable under chemical reducing conditions.

Graphite-ceramic rf Faraday-thermal shield and plasma limiter

Science.gov (United States)

Hwang, D.L.Q.; Hosea, J.C.

1983-05-05

The present invention is directed to a brazing procedure for joining a ceramic or glass material (e.g., Al/sub 2/O/sub 3/ or Macor) to graphite. In particular, the present invention is directed to a novel brazing procedure for the production of a brazed ceramic graphite product useful as a Faraday shield. The brazed ceramic graphite Faraday shield of the present invention may be used in Magnetic Fusion Devices (e.g., Princeton Large Torus Tokamak) or other high temperature resistant apparatus.

Initial Comparison of Baseline Physical and Mechanical Properties for the VHTR Candidate Graphite Grades

Energy Technology Data Exchange (ETDEWEB)

Carroll, Mark C. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Program

2014-09-01

High-purity graphite is the core structural material of choice in the Very High Temperature Reactor (VHTR) design, a graphite-moderated, helium-cooled configuration capable of producing thermal energy for power generation as well as process heat for industrial applications that require temperatures higher than the outlet temperatures of present nuclear reactors. The Baseline Graphite Characterization Program is establishing accurate as-manufactured mechanical and physical property distributions in nuclear-grade graphites by providing comprehensive data that captures the level of variation in measured values. In addition to providing a thorough comparison between these values in different graphite grades, the program is also carefully tracking individual specimen source, position, and orientation information in order to provide comparisons both in specific properties and in the associated variability between different lots, different billets, and different positions from within a single billet. This report is a preliminary comparison between each of the grades of graphite that are considered “candidate” grades from four major international graphite producers. These particular grades (NBG-18, NBG-17, PCEA, IG-110, and 2114) are the major focus of the evaluations presently underway on irradiated graphite properties through the series of Advanced Graphite Creep (AGC) experiments. NBG-18, a medium-grain pitch coke graphite from SGL from which billets are formed via vibration molding, was the favored structural material in the pebble-bed configuration. NBG-17 graphite from SGL is essentially NBG-18 with the grain size reduced by a factor of two. PCEA, petroleum coke graphite from GrafTech with a similar grain size to NBG-17, is formed via an extrusion process and was initially considered the favored grade for the prismatic layout. IG-110 and 2114, from Toyo Tanso and Mersen (formerly Carbone Lorraine), respectively, are fine-grain grades produced via an isomolding

Graphite Recycling from Spent Lithium-Ion Batteries.

Science.gov (United States)

Rothermel, Sergej; Evertz, Marco; Kasnatscheew, Johannes; Qi, Xin; Grützke, Martin; Winter, Martin; Nowak, Sascha

2016-12-20

The present work reports on challenges in utilization of spent lithium-ion batteries (LIBs)-an increasingly important aspect associated with a significantly rising demand for electric vehicles (EVs). In this context, the feasibility of anode recycling in combination with three different electrolyte extraction concepts is investigated. The first method is based on a thermal treatment of graphite without electrolyte recovery. The second method additionally utilizes a subcritical carbon-dioxide (subcritical CO 2 )-assisted electrolyte extraction prior to thermal treatment. And the final investigated approach uses supercritical carbon dioxide (scCO 2 ) as extractant, subsequently followed by the thermal treatment. It is demonstrated that the best performance of recycled graphite anodes can be achieved when electrolyte extraction is performed using subcritical CO 2 . Comparative studies reveal that, in the best case, the electrochemical performance of recycled graphite exceeds the benchmark consisting of a newly synthesized graphite anode. As essential efforts towards electrolyte extraction and cathode recycling have been made in the past, the electrochemical behavior of recycled graphite, demonstrating the best performance, is investigated in combination with a recycled LiNi 1/3 Co 1/3 Mn 1/3 O 2 cathode. © 2016 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Structure and nanotribology of thermally deposited gold nanoparticles on graphite

Energy Technology Data Exchange (ETDEWEB)

Cihan, Ebru [UNAM - Institute of Materials Science and Nanotechnology, Bilkent University, Ankara 06800 (Turkey); Özoğul, Alper [Department of Mechanical Engineering, Bilkent University, Ankara 06800 (Turkey); Baykara, Mehmet Z., E-mail: mehmet.baykara@bilkent.edu.tr [UNAM - Institute of Materials Science and Nanotechnology, Bilkent University, Ankara 06800 (Turkey); Department of Mechanical Engineering, Bilkent University, Ankara 06800 (Turkey)

2015-11-01

Graphical abstract: - Highlights: • Structure and tribology of thermally deposited AuNPs on HOPG have been studied. • Well-faceted, hexagonal AuNPs are formed on HOPG upon post-deposition annealing. • The crystalline character of the AuNPs is confirmed via TEM measurements. • AFM measurements reveal a “2/3” power law dependence of friction on load on AuNPs. • Friction forces at AuNP edges evolve linearly with increasing height and load. - Abstract: We present experiments involving the structural and frictional characterization of gold nanoparticles (AuNP) thermally deposited on highly oriented pyrolytic graphite (HOPG). The effect of thermal deposition amount, as well as post-deposition annealing on the morphology and distribution of gold on HOPG is studied via scanning electron microscopy (SEM) measurements, while transmission electron microscopy (TEM) is utilized to confirm the crystalline character of the nanoparticles. Lateral force measurements conducted via atomic force microscopy (AFM) under ambient conditions are employed to investigate the nanotribological properties of the gold nanoparticles as a function of normal load. Finally, the increase in lateral force experienced at the edges of the nanoparticles is studied as a function of normal load, as well as nanoparticle height. As a whole, our results constitute a comprehensive structural and frictional characterization of the AuNP/HOPG material system, forming the basis for nanotribology experiments involving the lateral manipulation of thermally deposited AuNPs on HOPG via AFM under ambient conditions.

Structure and nanotribology of thermally deposited gold nanoparticles on graphite

International Nuclear Information System (INIS)

Cihan, Ebru; Özoğul, Alper; Baykara, Mehmet Z.

2015-01-01

Graphical abstract: - Highlights: • Structure and tribology of thermally deposited AuNPs on HOPG have been studied. • Well-faceted, hexagonal AuNPs are formed on HOPG upon post-deposition annealing. • The crystalline character of the AuNPs is confirmed via TEM measurements. • AFM measurements reveal a “2/3” power law dependence of friction on load on AuNPs. • Friction forces at AuNP edges evolve linearly with increasing height and load. - Abstract: We present experiments involving the structural and frictional characterization of gold nanoparticles (AuNP) thermally deposited on highly oriented pyrolytic graphite (HOPG). The effect of thermal deposition amount, as well as post-deposition annealing on the morphology and distribution of gold on HOPG is studied via scanning electron microscopy (SEM) measurements, while transmission electron microscopy (TEM) is utilized to confirm the crystalline character of the nanoparticles. Lateral force measurements conducted via atomic force microscopy (AFM) under ambient conditions are employed to investigate the nanotribological properties of the gold nanoparticles as a function of normal load. Finally, the increase in lateral force experienced at the edges of the nanoparticles is studied as a function of normal load, as well as nanoparticle height. As a whole, our results constitute a comprehensive structural and frictional characterization of the AuNP/HOPG material system, forming the basis for nanotribology experiments involving the lateral manipulation of thermally deposited AuNPs on HOPG via AFM under ambient conditions.

Thermal Conductivity of a Monolayer of Exfoliated Graphite Nanoplatelets Prepared by Liquid-Liquid Interfacial Self-Assembly

Directory of Open Access Journals (Sweden)

Jinglei Xiang

2010-01-01

Full Text Available A monolayer film composed of exfoliated graphite nanoplatelets (xGnPs was extracted from a chloroform-water interface and supported on a glass substrate. The nanoplatelets are interconnected at the edges without overlapping forming a very densely packed structure with uniform thickness. Micro-Raman spectroscopy with a 50 mW 532 nm laser generating heat at the center of a xGnP sample was used to probe the thermal conductivity of the xGnP monolayer at different power levels. The Raman G peak shift of graphite was used to record the local temperature rise in the monolayer. The cross-sectional area of heat conduction is determined by the thickness of individual nanoplatelets. A UV-Vis spectrometer was used to measure the absorption of light by the monolayer. Depending on the interface density, the thermal conductivities are around 380 W/m K and 290 W/m K for monolayers with average particle size of 10 μm and 5 μm, respectively.

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

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)

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)

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...

Development of an apparatus for measuring the thermal conductivity of irradiated or non-irradiated graphite

International Nuclear Information System (INIS)

Bocquet, M.; Micaud, G.

1962-01-01

An apparatus was developed for measuring the thermal conductivity coefficient K of irradiated or non-irradiated graphite. The measurement of K at around room temperature with an accuracy of about 6% is possible. The study specimen is placed in a vacuum between a hot and a cold source which create a temperature gradient ΔΘ/ Δx in the steady state. The amount of heat transferred, Q, is deduced from the electrical power dissipated at the hot source, after allowing for heat losses. The thermal conductivity coefficient is defined as: K = Q/S. Δx/ΔΘ, S being the cross section of the sample. Systematic studies have made it possible to determine the mean values of the thermal conductivity. (authors) [fr

Thermal conductivity of partially graphitized biocarbon obtained by carbonization of medium-density fiberboard in the presence of a Ni-based catalyst

Science.gov (United States)

Orlova, T. S.; Parfen'eva, L. S.; Smirnov, B. I.; Gutierrez-Pardo, A.; Ramirez-Rico, J.

2016-01-01

The thermal conductivity k and resistivity ρ of biocarbon matrices, prepared by carbonizing medium-density fiberboard at T carb = 850 and 1500°C in the presence of a Ni-based catalyst (samples MDF-C( Ni)) and without a catalyst (samples MDF-C), have been measured for the first time in the temperature range of 5-300 K. X-ray diffraction analysis has revealed that the bulk graphite phase arises only at T carb = 1500°C. It has been shown that the temperature dependences of the thermal conductivity of samples MDFC- 850 and MDF-C-850(Ni) in the range of 80-300 K are to each other and follow the law of k( T) ˜ T 1.65, but the use of the Ni-catalyst leads to an increase in the thermal conductivity by a factor of approximately 1.5, due to the formation of a greater fraction of the nanocrystalline phase in the presence of the Ni-catalyst at T carb = 850°C. In biocarbon MDF-C-1500 prepared without a catalyst, the dependence is k( T) ˜ T 1.65, and it is controlled by the nanocrystalline phase. In MDF-C-1500(Ni), the bulk graphite phase formed increases the thermal conductivity by a factor of 1.5-2 compared to the thermal conductivity of MDF-C-1500 in the entire temperature range of 5-300 K; k( T = 300 K) reaches the values of ˜10 W m-1 K-1, characteristic of biocarbon obtained without a catalyst only at high temperatures of T carb = 2400°C. It has been shown that MDF-C-1500(Ni) in the temperature range of 40‒300 K is characterized by the dependence, k( T) ˜ T 1.3, which can be described in terms of the model of partially graphitized biocarbon as a composite of an amorphous matrix with spherical inclusions of the graphite phase.

The role of graphite morphology and matrix structure on low ...

Indian Academy of Sciences (India)

R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

Thermal cycling resistance; graphite morphology; grey cast iron; austempered ductile iron; compacted/vermicular graphite iron; matrix decompo- sition. 1. Introduction. When a material is subjected to a temperature gradient, it tends to expand differentially. During this process, thermal stresses are induced. The source of ...

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.

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

A 3-D Thermal Analysis of the HANARO Cold Neutron Moderator Cell

International Nuclear Information System (INIS)

Han, Gee Y.; Kim, Heo Nil

2007-01-01

Fundamental studies on a thermal analysis of a cryogenic system such as a cold neutron source (CNS) have increased significantly for a successful CNS design in cold neutron research during recent years. A three-dimensional (3-D) thermal analysis model for the HANARO CNS was developed and used to accurately predict a temperature distribution between the hydrogen inside and the entire inner and outer surfaces of a moderator cell, whose moderator and cell walls are heated differently, under a steady-state operating condition by using the HEATING 7 code. The objective of this study is primarily to predict a temperature distribution through a heat flow in a cold neutron moderator cell heated from a nuclear heating and cooled by a cryogenic coolant. This paper presents satisfactory results of a steady-state temperature distribution in a cryogenic moderator cell. They are used to support the thermal stress analysis of the moderator cell walls and to provide a safe operation for the HANARO CNS facility

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

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)

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Study of graphite reactivity worth on well-defined cores assembled on LR-0 reactor

International Nuclear Information System (INIS)

Košťál, Michal; Rypar, Vojtěch; Milčák, Ján; Juříček, Vlastimil; Losa, Evžen; Forget, Benoit; Harper, Sterling

2016-01-01

Highlights: • A light water critical facility for graphite reactivity worth measurements. • Comparison of calculated and measured k eff . • Effect of graphite description on k eff . - Abstract: Graphite is an often-used moderating material on the basis of its good moderating power and very low absorption cross section. This small absorption cross section permits the use of natural or low-enriched uranium in graphite moderated reactors. Graphite is now being considered as the moderator for Fluoride-salt-cooled High Temperature Reactors (FHR). The critical moderator level was measured for various graphite block configurations in an experimental dry assembly of the LR-0 reactor. Comparisons with experiments were performed between Monte Carlo simulation tools for which satisfactory agreement was obtained with the exception of some systematic discrepancies. The larger discrepancies were observed when using the ENDF/B-VII.0 library. To decrease the uncertainties, based on conservative assumptions, relative comparisons were done. The results provided by the different nuclear data libraries are within 3 sigma interval of experimental uncertainties. It has been determined that differences between the results of calculations are caused by variations in the (n,n), (n,n′), (n,g) reactions and also by various angular distributions, while the (n,g) cross section variations play only a minor role for these configurations.

Diffusion of graphite. The effect of cylindrical canals; Longueur de diffusion du graphite effet des canaux cylindriques

Energy Technology Data Exchange (ETDEWEB)

Carle, R; Clouet d' Orval, C; Martelly, J; Mazancourt, T de; Sagot, M; Lattes, R; Teste du Bailler, A [Commissariat a l' Energie Atomique, Dir. Industrielle, Saclay (France). Centre d' Etudes Nucleaires; Robert, C [Ecole Normale Superieure, 75 - Paris (France)

1957-07-01

Experiments on thermal neutron diffusion in the graphite used as moderator in the pile G1 have been carried out. The object of these experiments is to determine: - the intrinsic quality of this graphite, characterised by its diffusion length L or its Laplacian 1/L{sup 2} - the effect of the canals, which modifies anisotropically the macroscopic diffusion equation and is characterized by two principal diffusion regions (or two principal Laplacian), valid respectively for the diffusion in the direction of the canals and in a perpendicular direction. In order to determine them two experiments are necessary, in which the second derivatives of the flux in relation to the space coordinates are very different. These experiments form the object of the first two parts. Part 1: Diffusion along the axis of a flux coming from the pile source, and limited radially by a quasi cylindrical screen of cadmium bars. This screen, or Faraday cage is designed to give to the thermal flux produced the same radius of extrapolation to zero as that of the pile source. The determination of L (with the graphite full) has been made under the same conditions. The measurements have been interpreted in two ways. The influence of the brackets holding the detectors is discussed. Part 2: Radial diffusion in the graphite surrounding the 'long' cylindrical pile. This is well described by a sum of Bessel functions. Part 3: Results (valid for d = 1.61 t = 17 deg. C). For the graphite without cavity L = 52.7 {+-} 0.4 cm. The effect of the canals on the diffusion area and its anisotropy are in excellent agreement with the theory of Behrens: L(parallel) = 64.6 cm and L(perpendicular) 62.2 cm. Appendix: Theory of the Faraday cage. (author) [French] Des experiences de diffusion des neutrons thermiques dans le graphite constituant le moderateur de la pile G1 ont ete effectuees. Elles ont pour objet de determiner: - la qualite intrinseque de ce graphite, caracterisee par sa longueur de diffusion L ou son

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

Paraffin/expanded graphite phase change composites with enhanced thermal conductivity prepared by implanted β-SiC nanowires with chemical vapor deposition method

Science.gov (United States)

Yin, Zhaoyu; Zhang, Xiaoguang; Huang, Zhaohui; Liu, Silin; Zhang, Weiyi; Liu, Yan'gai; Wu, Xiaowen; Fang, Minghao; Min, Xin

2018-02-01

Expanded graphite/β-SiC nanowires composites (ESNC) were prepared through chemical vapor deposition, and paraffin/expanded graphite/β-SiC nanowires composites (PESNC) were made through vacuum impregnation to overcome liquid leakage during phase transition and enhance the thermal conductivity of paraffin. Fourier transform infrared spectroscopy showed no chemical interactions between the paraffin and ESNC. Differential scanning calorimetry estimated the temperature and latent heat of PESNC during melting to 45.73 °C and 124.31 J g-1, respectively. The respective values of these quantities during freezing were recorded as 48.93 °C and 124.14 J g-1. The thermal conductivity of PESNC was estimated to 0.75 W mK-1, which was 3.26-folds that of pure paraffin (0.23 W mK-1). PESNC perfectly maintained its phase transition after 200 melting-freezing cycles. The resulting ideal thermal conductivity, good chemical stability, thermal properties and thermal reliability of PESNC are promising for use in energy efficient buildings and solar energy systems.

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

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

Synthesis of Cu-coated Graphite Powders Using a Chemical Reaction Process

Energy Technology Data Exchange (ETDEWEB)

Jang, Jun-Ho; Park, Hyun-Kuk; Oh, Ik-Hyun [Korea Institute of Industrial Technology (KITECH), Gwangju (Korea, Republic of); Lim, Jae-Won [Chonbuk National University, Jeonju (Korea, Republic of)

2017-05-15

In this paper, Cu-coated graphite powders for a low thermal expansion coefficient and a high thermal conductivity are fabricated using a chemical reaction process. The Cu particles adhere to the irregular graphite powders and they homogeneously disperse in the graphite matrix. Cu-coated graphite powders are coarser at approximately 3-4 μm than the initial graphite powders; furthermore, their XRD patterns exhibit a low intensity in the oxide peak with low Zn powder content. For the passivation powders, the transposition solvent content has low values, and the XRD pattern of the oxide peaks is almost non-existent, but the high transposition solvent content does not exhibit a difference to the non-passivation treated powders.

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

Science.gov (United States)

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

2013-06-25

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

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.)

Condensation nuclear power plants with water-cooled graphite-moderated channel type reactors and advances in their development

International Nuclear Information System (INIS)

Boldyrev, V.M.; Mikhaj, V.I.

1985-01-01

Consideration is being given to results of technical and economical investigations of advisability of increasing unit power by elevating steam generating capacity as a result of inserting numerous of stereotype sectional structural elements of the reactor with similar thermodynamic parameters. It is concluded that construction of power units of condensation nuclear power plants with water-cooled graphite-moderated channel type reactors of 2400-3200 MWe and higher unit power capacity represents the real method for sharp growth of efficiency and labour productivity in power industry. It can also provide the required increase of the rate of putting electrogenerating powers into operation

Changes in Effective Thermal Conductivity During the Carbothermic Reduction of Magnetite Using Graphite

Science.gov (United States)

Kiamehr, Saeed; Ahmed, Hesham; Viswanathan, Nurni; Seetharaman, Seshadri

2017-06-01

Knowledge of the effective thermal diffusivity changes of systems undergoing reactions where heat transfer plays an important role in the reaction kinetics is essential for process understanding and control. Carbothermic reduction process of magnetite containing composites is a typical example of such systems. The reduction process in this case is highly endothermic and hence, the overall rate of the reaction is greatly influenced by the heat transfer through composite compact. Using Laser-Flash method, the change of effective thermal diffusivity of magnetite-graphite composite pellet was monitored in the dynamic mode over a pre-defined thermal cycle (heating at the rate of 7 K/min to 1423 K (1150 °C), holding the sample for 270 minutes at this temperature and then cooling it down to the room temperature at the same rate as heating). These measurements were supplemented by Thermogravimetric Analysis under comparable experimental conditions as well as quenching tests of the samples in order to combine the impact of various factors such as sample dilatations and changes in apparent density on the progress of the reaction. The present results show that monitoring thermal diffusivity changes during the course of reduction would be a very useful tool in a total understanding of the underlying physicochemical phenomena. At the end, effort is made to estimate the apparent thermal conductivity values based on the measured thermal diffusivity and dilatations.

Thermal design and validation of radiation detector for the ChubuSat-2 micro-satellite with high-thermal-conductive graphite sheets

Science.gov (United States)

Park, Daeil; Miyata, Kikuko; Nagano, Hosei

2017-07-01

This paper describes thermal design of the radiation detector (RD) for the ChubuSat-2 with the use of high-thermal-conductive materials. ChubuSat-2 satellite is a 50-kg-class micro-satellite joint development with Nagoya University and aerospace companies. The main mission equipment of ChubuSat-2 is a RD to observe neutrons and gamma rays. However, the thermal design of the RD encounters a serious problem, such as no heater for RD and electric circuit alignment constrain. To solve this issue, the RD needs a new thermal design and thermal control for successful space missions. This paper proposes high-thermal-conductive graphite sheets to be used as a flexible radiator fin for the RD. Before the fabrication of the device, the optimal thickness and surface area for the flexible radiator fin were determined by thermal analysis. Consequently, the surface area of flexible radiator fin was determined to be 8.6×104 mm2. To verify the effects of the flexible radiator fin, we constructed a verification model and analyzed the temperature distributions in the RD. Also, the thermal vacuum test was performed using a thermal vacuum chamber, which was evacuated at a pressure of around 10-4 Pa, and its internal temperature was cooled at -80 °C by using a refrigerant. As a result, it has been demonstrated that the flexible radiator fin is effective. And the thermal vacuum test results are presented good correlation with the analysis results.

A study of the relationship between microstructure and oxidation effects in nuclear graphite at very high temperatures

Science.gov (United States)

Lo, I.-Hsuan; Tzelepi, Athanasia; Patterson, Eann A.; Yeh, Tsung-Kuang

2018-04-01

Graphite is used in the cores of gas-cooled reactors as both the neutron moderator and a structural material, and traditional and novel graphite materials are being studied worldwide for applications in Generation IV reactors. In this study, the oxidation characteristics of petroleum-based IG-110 and pitch-based IG-430 graphite pellets in helium and air environments at temperatures ranging from 700 to 1600 °C were investigated. The oxidation rates and activation energies were determined based on mass loss measurements in a series of oxidation tests. The surface morphology was characterized by scanning electron microscopy. Although the thermal oxidation mechanism was previously considered to be the same for all temperatures higher than 1000 °C, the significant increases in oxidation rate observed at very high temperatures suggest that the oxidation behavior of the selected graphite materials at temperatures higher than 1200 °C is different. This work demonstrates that changes in surface morphology and in oxidation rate of the filler particles in the graphite materials are more prominent at temperatures above 1200 °C. Furthermore, possible intrinsic factors contributing to the oxidation of the two graphite materials at different temperature ranges are discussed taking account of the dominant role played by temperature.

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

Preliminary thermal-hydraulic and structural strength analyses for pre-moderator of cold moderator

International Nuclear Information System (INIS)

Aso, Tomokazu; Kaminaga, Masanori; Terada, Atsuhiko; Hino, Ryutaro

2001-08-01

A light-water cooled pre-moderator with a thin-walled structure made of aluminum alloy is installed around a liquid hydrogen moderator in order to enhance the neutron performance of a MW-scale spallation target system which is being developed in the Japan Atomic Energy Research Institute (JAERI). Since the pre-moderator is needed to be located close to a target working as a neutron source, it is indispensable to remove nuclear heat deposition in the pre-moderator effectively by means of smooth water flow without flow stagnation. Also, the structural integrity of the thin-walled structure should be kept against the water pressure. Preliminary thermal-hydraulic analytical results showed that the water temperature rise could be suppressed less than 1degC while keeping the smooth water flow, which would assure the expected neutron performance. As for the structural integrity, several measures to meet allowable stress conditions of aluminum alloy were proposed on the basis of the preliminary structural strength analyses. (author)

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)

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)

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

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)

Some physical methods for study of irradiation effects in graphite; Quelques procedes physiques pour etudier les effets de l'irradiation du graphite

Energy Technology Data Exchange (ETDEWEB)

Mayer, G; Lecomte, M; Mattmuller, R [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1959-07-01

A calibration method for a classical apparatus for differential thermal analysis is described in detail. This method achieves a relative precision of 5 per cent in the measurement of the internal energy release accompanying the annealing of irradiated graphites. Elastic constants of graphites are obtained from the frequencies of the longitudinal modes of vibration; procedures for excitation and detection of these vibrations at any temperature between -190 deg. C and +1500 deg. C are described. A procedure for obtaining easily measured deformations of graphites after relatively little irradiation with thermal neutrons is discussed. An application of this method to the study of the thermal annealing of elongation caused by displaced atoms is indicated. (author) [French] On decrit en detail une methode d'etalonnage pour un appareil classique d'analyse thermique differentielle. Cette methode permet de mesurer avec une precision relative de 5% la liberation d'energie interne qui accompagne le 'recuit' des graphites irradies. On deduit les constantes elastiques des graphites des frequences des vibrations longitudinales et on decrit les procedes pour exciter et detecter ces vibrations a toutes les temperatures comprises entre -190 deg. C et + 1500 deg. C. On discute un procede pour obtenir une des deformations de graphites facilement mesurables apres une irradiation relativement faible a l'aide de neutrons thermiques. Une application de cette methode a l'etude du 'recuit' thermique de l'elongation causee par les atomes deplaces est indiquee. (auteur)

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

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.

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

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

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

From Core to Capture: Graphite Management by Gasification and Carbon Capture & Storage (CCS)

International Nuclear Information System (INIS)

Goodwin, J.; Bradbury, D.; Black, S.; Tomlinson, T.; Livesey, B.; Robinson, J.; Lindberg, M.; Newton, C.; Jones, A.; Wickham, A.

2016-01-01

Radioactive graphite waste arises principally from the moderators of graphite/gas-cooled reactors at the end of life of the reactors. Commercial power producing reactors (for example, Magnox, AGR and RBMK) have graphite moderators, each containing several thousand tonnes of graphite, with the UK having the largest inventory of over 90,000 tonnes. Additionally, there are smaller quantities of graphite arising from other sources such as fuel element components. The current long term strategy for management of reactor graphite in the UK is for these wastes to be conditioned for disposal followed by transfer to a geological disposal facility (GDF). With this baseline position, these wastes will account for about 30% of the ILW inventory in a GDF. As the volume of the graphite waste is so large, it is not currently economic to retrieve and process the graphite in advance of the availability of a geological disposal facility. Recent work by the NDA has ascribed a much smaller “incremental” volume of 2% due to graphite, calculated on the basis that the GDF has to be a certain size anyway in order to dissipate the decay heat from high level waste

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)

Application of INAA for chemical quality control analysis of C-C composite and high purity graphite by determining trace elemental concentrations

International Nuclear Information System (INIS)

Shinde, Amol D.; Reddy, A.V.R.; Acharya, R.; Venugopalan, Ramani

2015-01-01

Carbon based materials like graphite and C-C composites are used for various scientific and technological applications. Owing to its low neutron capture cross section and good moderating properties, graphite is used as a moderator or reflector in nuclear reactors. For high temperature reactors like CHTR, graphite and C-C composites are proposed as structural materials. Studies are in progress to use C-C composites as prospective candidate instead of graphite due to their excellent mechanical and thermal properties. The advantage of carbon-carbon composite is that the microstructure and the properties can be tailor made. Impurities like rare earth elements and neutron poisons which have high neutron absorption cross section and elements whose activation products of have longer half-lives like 60 Co (5.27 y), 65 Zn (244.3 d) and 59 Fe (44.5 d) are not desired in structural materials. For chemical quality control (CQC) it is necessary to evaluate accurately the impurity concentrations using a suitable non-destructive analytical technique. In the present work, two carbon/carbon composite samples and two high purity graphite samples were analyzed by Instrumental Neutron Activation Analysis (INAA) using high-flux reactor neutrons. Samples, sealed in Al foil, were irradiated in tray-rod position of Dhruva reactor, BARC at a neutron flux of ∼ 5 x 10 13 cm -2 s -1 . Radioactive assay was carried out using high resolution gamma ray spectrometry using 40% HPGe detector

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

Radiogenic lead with dominant content of {sup 208}Pb: New coolant and neutron moderator for innovative nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Shmelev, A. N.; Kulikov, G. G.; Kryuchkov, E. F.; Apse, V. A.; Kulikov, E. G. [National Research Nuclear Univ. MEPhI, Kashirskoe shosse, 31, 115409, Moscow (Russian Federation)

2012-07-01

The advantages of radiogenic lead with dominant content of {sup 208}Pb as a reactor coolant with respect to natural lead are caused by unique nuclear properties of {sup 208}Pb which is a double-magic nucleus with closed proton and neutron shells. This results in significantly lower micro cross section and resonance integral of radiative neutron capture by {sup 208}Pb than those for numerous light neutron moderators. The extremely weak ability of {sup 208}Pb to absorb neutrons results in the following effects. Firstly, neutron moderating factor (ratio of scattering to capture cross sections) is larger than that for graphite and light water. Secondly, age and diffusion length of thermal neutrons are larger than those for graphite, light and heavy water. Thirdly, neutron lifetime in {sup 208}Pb is comparable with that for graphite, beryllium and heavy water what could be important for safe reactor operation. The paper presents some results obtained in neutronics and thermal-hydraulics evaluations of the benefits from the use of radiogenic lead with dominant content of {sup 208}Pb instead of natural lead as a coolant of fast breeder reactors. The paper demonstrates that substitution of radiogenic lead for natural lead can offer the following benefits for operation of fast breeder reactors. Firstly, improvement of the reactor safety thanks to the better values of coolant temperature reactivity coefficient and, secondly, improvement of some thermal-hydraulic reactor parameters. Radiogenic lead can be extracted from thorium sludge without isotope separation as {sup 208}Pb is a final isotope in the decay chain of {sup 232}Th. (authors)

Derivation of a radionuclide inventory for irradiated graphite-chlorine-36 inventory determination

International Nuclear Information System (INIS)

Brown, F.J.; Palmer, J.D.; Wood, P.

2001-01-01

The irradiation of materials in nuclear reactors results in neutron activation of component elements. Irradiated graphite wastes arise from their use in UK gas-cooled research and commercial reactor cores, and in fuel element components, where the graphite has acted as the neutron moderator. During irradiation the residual chlorine, which was used to purify the graphite during manufacture, is activated to chlorine-36. This isotope is long-lived and poorly retarded by geological barriers, and may therefore be a key radionuclide with respect to post-closure disposal facilities performance. United Kingdom Nirex Limited, currently responsible for the development of a disposal route for intermediate-level radioactive wastes in the UK, carried out a major research programme to support an overall assessment of the chlorine-36 activity of all wastes including graphite reactor components. The various UK gas cooled reactors reactors have used a range of graphite components made from diverse graphite types; this has necessitated a systematic programme to cover the wide range of graphite and production processes. The programme consisted of: precursor measurements - on the surface and/or bulk of representative samples of relevant materials, using specially developed methods; transfer studies - to quantify the potential for transfer of Cl-36 into and between waste streams during irradiation of graphite; theoretical assessments - to support the calculational methodology; actual measurements - to confirm the modelling. For graphite, a total of 458 measurements on samples from 57 batches were performed, to provide a detailed understanding of the composition of nuclear graphite. The work has resulted in the generation of probability density functions (PDF) for the mean chlorine concentration of three classes of graphite: fuel element graphite; Magnox moderator and reflector graphite and AGR reflector graphite; AGR moderator graphite. Transfer studies have shown that a significant

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)

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)

The development of a graphite to copper joint for injector diagnostics

International Nuclear Information System (INIS)

Baker, D.

1975-09-01

A successful method of joining soft graphite to copper, to achieve a high thermal conductivity joint, is described, together with initial attempts and the difficulties experienced. The graphite, after an initial treatment, was flame sprayed with copper. The main copper-to-graphite joint was of silver solder. (U.K.)

Comparisons of thermal SCWR assembly designs by in- or inter-assembly moderation

International Nuclear Information System (INIS)

Zhang, P.; Wang, K.

2010-01-01

As one of the six GEN-IV reactor systems and the only one with water as coolant, SCWR is thought to be the most hopeful future nuclear energy system. Many designs have already been proposed worldwide. For thermal SCWR designs, a key consideration factor is how to provide the dedicated moderation, which leads to various designs with water as moderator. It is much like BWR, where under-moderation is an important issue and central water rods are adopted to account for it. Naturally, water rods can be adopted in SCWR assemblies too, but with counter-flow scheme to increase the outlet temperature and thermal efficiency which is unlike the co-flow scheme in BWR. This type can be seen in the American, Japanese and European HPLWR designs. As the other option, moderation could be provided with inter-assembly gap like CANDU-SCWR design. It is worthy of review of these designs for better understanding the water moderating effects and putting forward any new designs. (author)

A microstructural study of dynamic crack propagation in nuclear graphites

International Nuclear Information System (INIS)

Burchell, T.D.; McEnaney, B.; Tucker, M.O.; Rose, A.P.G.

1986-01-01

This paper reports a new microstructural study of dynamic crack propagation in three nuclear graphites: (i) PGA, the moderator material in UK Magnox reactors; (ii) IMl-24, the moderator material in UK Advanced gas cooled reactors (AGR); and (iii) a pitch coke graphite, which is used in the fabrication of AGR fuel sleeves. The fracture mechanisms in nuclear graphites are initiated by microcrack formation at low stresses. Typically, microcracks form in regions of well-aligned binder or at favourably-oriented pores, where stress is concentrated. With increasing applied loads, microcracks propagate taking advantage of easy cleavage paths or linking with pores. Eventually, coalescence of such cracks and inherent porosity produces a crack of critical length for fast fracture. (orig./MM)

Ion irradiation of {sup 37}Cl implanted nuclear graphite: Effect of the energy deposition on the chlorine behavior and consequences for the mobility of {sup 36}Cl in irradiated graphite

Energy Technology Data Exchange (ETDEWEB)

Toulhoat, N., E-mail: nelly.toulhoat@univ-lyon1.fr [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); CEA/DEN, Centre de Saclay (France); Moncoffre, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (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); 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); Sainsot, P. [Université de Lyon, Université Lyon 1, LaMCoS, INSA-Lyon, CNRS UMR5259 (France); Rouzaud, J.-N.; Deldicque, D. [Laboratoire de Géologie de l’Ecole Normale Supérieure (ENS), Paris, UMR CNRS-ENS 8538 (France)

2015-09-15

Graphite is used in many types of nuclear reactors due to its ability to slow down fast neutrons without capturing them. Whatever the reactor design, the irradiated graphite waste management has to be faced sooner or later regarding the production of long lived or dose determining radioactive species such as {sup 14}C, {sup 3}H or {sup 36}Cl. The first carbon dioxide cooled, graphite moderated nuclear reactors resulted in a huge quantity of irradiated graphite waste for which the management needs a previous assessment of the radioactive inventory and the radionuclide’s location and speciation. As the detection limits of usual spectroscopic methods are generally not adequate to detect the low concentration levels (<1 ppm) of the radionuclides, we used an indirect approach based on the implantation of {sup 37}Cl, to simulate the presence of {sup 36}Cl. Our previous studies show that temperature is one of the main factors to be considered regarding the structural evolution of nuclear graphite and chlorine mobility during reactor operation. However, thermal release of chlorine cannot be solely responsible for the depletion of the {sup 36}Cl inventory. We propose in this paper to study the impact of irradiation and its synergetic effects with temperature on chlorine release. Indeed, the collision of the impinging neutrons with the graphite matrix carbon atoms induces mainly ballistic collisions. However, a small part of the recoil carbon atom energy is also transferred to the lattice through electronic excitation. This paper aims at elucidating the effects of the different irradiation regimes (ballistic and electronic) using ion irradiation, on the mobility of implanted {sup 37}Cl, taking into account the initial disorder level of the nuclear graphite.

Mechanical and thermophysical properties of graphite/polyimide composite materials

Science.gov (United States)

Rummler, D. R.; Clark, R. K.

1979-01-01

An on-going program to characterize advanced composites for up to 50,000 hours of exposure to simulated supersonic cruise environments is summarized. Results are presented for up to 25,000 hours of thermal exposure and 10,000 hours of flight simulation at temperatures up to 560K (550 F) with emphasis on HTS/710 graphite/polyimide composite material. Results to date indicate that the maximum use temperature for HTS/710 may be reduced to 505K (450 F) for long-time (1000 hours) application such as the supersonic transport. Preliminary thermophysical properties data for HTS/PMR15 graphite/polyimide were generated. These data include thermal conductivity, thermal expansion, and specific heat from 115K (-252 F) to 590K (600 F) and emittance at room temperature and 590K (600 F). The purpose in generating these data was to validate use of state-of-the-art property measurement methods for advanced graphite fiber reinforced resin matrix composites. Based on results to this point, thermal expansion measurements for composites are most difficult to perform. A high degree of caution in conducting thermal expansion tests and analyzing results is required to produce reliable data.

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

Thermophysical properties and microstructure of graphite flake/copper composites processed by electroless copper coating

International Nuclear Information System (INIS)

Liu, Qian; He, Xin-Bo; Ren, Shu-Bin; Zhang, Chen; Ting-Ting, Liu; Qu, Xuan-Hui

2014-01-01

Highlights: • GF–copper composites were fabricated using a sparking plasma sintering, which involves coating GF with copper, using electroless plating technique. • The oriented graphite flake distributed homogeneously in matrix. • With the increase of flake graphite from 44 to 71 vol.%, the basal plane thermal conductivity of composites increases from 445 to 565 W m −1 K −1 and the thermal expansion of composites decreases from 8.1 to 5.0. • The obtained composites are suitable for electronic packaging materials. -- Abstract: This study focuses on the fabrication of thermal management material for power electronics applications using graphite flake reinforced copper composites. The manufacturing route involved electroless plating of copper on the graphite flake and further spark plasma sintering of composite powders. The relative density of the composites with 44–71 vol.% flakes achieved up to 98%. Measured thermal conductivities and coefficients of thermal expansion of composites ranged from 455–565 W m −1 K −1 and 8 to 5 ppm K −1 , respectively. Obtained graphite flake–copper composites exhibit excellent thermophysical properties to meet the heat dispersion and matching requirements of power electronic devices to the packaging materials

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

International Nuclear Information System (INIS)

Johnson, P.A.V.

1985-09-01

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

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.)

Effect of Graphite on the Properties of Natural Rubber

Directory of Open Access Journals (Sweden)

Auda jabber Braihi

2016-09-01

Full Text Available Natural rubber-graphite composites (0, 1, 2, 3, 4 pphr graphite were prepared on a laboratory two-roll mill. Swelling measurements were used to evaluate the impacts of graphite on the properties of natural rubber. Swelling results showed that the volume fraction of natural rubber in the swollen gel, the interaction parameter, and the cross-link density decreased by increasing graphite loadings, while the average molecular weight of natural rubber between cross-links increased. Vulcanization results showed that only scorch time parameter increased with increasing graphite loadings, while other parameters (Max. torque, Min. torque, cure rate and cure rate index decreased. Both thermal and AC conductivities increased.

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)

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

Radiogenic Lead with Dominant Content of 208Pb: New Coolant and Neutron Moderator for Innovative Nuclear Facilities

Directory of Open Access Journals (Sweden)

A. N. Shmelev

2011-01-01

Full Text Available As a rule materials of small atomic weight (light and heavy water, graphite, and so on are used as neutron moderators and reflectors. A new very heavy atomic weight moderator is proposed—radiogenic lead consisting mainly of isotope 208Pb. It is characterized by extremely low neutron radiative capture cross-section (0.23 mbarn for thermal neutrons, i.e., less than that for graphite and deuterium and highest albedo of thermal neutrons. It is evaluated that the use of radiogenic lead makes it possible to slow down the chain fission reaction on prompt neutrons in a fast reactor. This can increase safety of the fast reactors and reduce as well requirements pertaining to the fuel fabrication technology. Radiogenic lead with high 208Pb content as a liquid-metal coolant of fast reactors helps to achieve a favorable (negative reactivity coefficient on coolant temperature. It is noteworthy that radiogenic lead with high 208Pb content may be extracted from thorium (as well as thorium-uranium ores without isotope separation. This has been confirmed experimentally by the investigations performed at San Paulo University, Brazil.

The generalized heavy free gas thermalization operator

International Nuclear Information System (INIS)

Pitcher, H.H.W.

1963-05-01

This paper gives an introduction to the generalized heavy free gas (Horowitz) thermalization operator, which is a device for simplifying thermalization calculations. Analytical, experimental,and numerical methods for determining the energy-dependent function f in the operator are discussed. The dependence of f on the nature of the moderator, its temperature, and the absorption of the system are investigated, using a program FOCS which derives f's from DSN (multigroup) thermal spectra; it is found that for most purposes the dependence on absorption is negligible except when Pu240 is present. The sensitivity of calculated spectra and reaction rate to changes in f is considered. f is given for graphite at 300 and 600 deg. K. (author)

Differences between young adults and elderly in thermal comfort, productivity and thermal physiology in response to a moderate temperature drift

DEFF Research Database (Denmark)

Schellen, Lisje; Lichtenbelt, Wouter van Marken; Loomans, Marcel

2010-01-01

thermal condition differ between young adults and elderly. There is a lack of studies that describe the effect of aging on thermal comfort and productivity during a moderate temperature drift. In this study, the effect of a moderate temperature drift on physiological responses, thermal comfort......Results from naturally ventilated buildings show that allowing the indoor temperature to drift does not necessarily result in thermal discomfort and may allow for a reduction in energy use. However, for stationary conditions, several studies indicate that the thermal neutral temperature and optimum......, temperature drift: first 4 h: +2 K/h, last 4 h: –2 K/h. The results indicate that thermal sensation of the elderly was, in general, 0.5 scale units lower in comparison with their younger counterparts. Furthermore, the elderly showed more distal vasoconstriction during both conditions. Nevertheless, TS...

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Dependence of strength on particle size in graphite

International Nuclear Information System (INIS)

Kennedy, E.P.; Kennedy, C.R.

The strength to particle size relationship for specially fabricated graphites has been demonstrated and rationalized using fracture mechanics. In the past, similar studies have yielded empirical data using only commercially available material. Thus, experimental verification of these relationships has been difficult. However, the graphites of this study were fabricated by controlling the particle size ranges for a series of isotropic graphites. All graphites that were evaluated had a constant 1.85 g/cm 3 density. Thus, particle size was the only variable. This study also considered the particle size effect on other physical properties; coefficient of thermal expansion (CTE), electrical resistivity, fracture strain, and Young's modulus

Improvement in transdermal drug delivery performance by graphite oxide/temperature-responsive hydrogel composites with micro heater

International Nuclear Information System (INIS)

Yun, Jumi; Lee, Dae Hoon; Im, Ji Sun; Kim, Hyung-Il

2012-01-01

Transdermal drug delivery system (TDDS) was prepared with temperature-responsive hydrogel. The graphite was oxidized and incorporated into hydrogel matrix to improve the thermal response of hydrogel. The micro heater was fabricated to control the temperature precisely by adopting a joule heating method. The drug in hydrogel was delivered through a hairless mouse skin by controlling temperature. The efficiency of drug delivery was improved obviously by incorporation of graphite oxide due to the excellent thermal conductivity and the increased interfacial affinity between graphite oxide and hydrogel matrix. The fabricated micro heater was effective in controlling the temperature over lower critical solution temperature of hydrogel precisely with a small voltage less than 1 V. The cell viability test on graphite oxide composite hydrogel showed enough safety for using as a transdermal drug delivery patch. The performance of TDDS could be improved noticeably based on temperature-responsive hydrogel, thermally conductive graphite oxide, and efficient micro heater. - Graphical abstract: The high-performance transdermal drug delivery system could be prepared by combining temperature-responsive hydrogel, thermally conductive graphite oxide with improved interfacial affinity, and efficient micro heater fabricated by a joule heating method. Highlights: ► High performance of transdermal drug delivery system with an easy control of voltage. ► Improved thermal response of hydrogel by graphite oxide incorporation. ► Efficient micro heater fabricated by a joule heating method.

Thermal desorption of formamide and methylamine from graphite and amorphous water ice surfaces

Science.gov (United States)

Chaabouni, H.; Diana, S.; Nguyen, T.; Dulieu, F.

2018-04-01

Context. Formamide (NH2CHO) and methylamine (CH3NH2) are known to be the most abundant amine-containing molecules in many astrophysical environments. The presence of these molecules in the gas phase may result from thermal desorption of interstellar ices. Aims: The aim of this work is to determine the values of the desorption energies of formamide and methylamine from analogues of interstellar dust grain surfaces and to understand their interaction with water ice. Methods: Temperature programmed desorption (TPD) experiments of formamide and methylamine ices were performed in the sub-monolayer and monolayer regimes on graphite (HOPG) and non-porous amorphous solid water (np-ASW) ice surfaces at temperatures 40-240 K. The desorption energy distributions of these two molecules were calculated from TPD measurements using a set of independent Polanyi-Wigner equations. Results: The maximum of the desorption of formamide from both graphite and ASW ice surfaces occurs at 176 K after the desorption of H2O molecules, whereas the desorption profile of methylamine depends strongly on the substrate. Solid methylamine starts to desorb below 100 K from the graphite surface. Its desorption from the water ice surface occurs after 120 K and stops during the water ice sublimation around 150 K. It continues to desorb from the graphite surface at temperatures higher than160 K. Conclusions: More than 95% of solid NH2CHO diffuses through the np-ASW ice surface towards the graphitic substrate and is released into the gas phase with a desorption energy distribution Edes = 7460-9380 K, which is measured with the best-fit pre-exponential factor A = 1018 s-1. However, the desorption energy distribution of methylamine from the np-ASW ice surface (Edes = 3850-8420 K) is measured with the best-fit pre-exponential factor A = 1012 s-1. A fraction of solid methylamine monolayer of roughly 0.15 diffuses through the water ice surface towards the HOPG substrate. This small amount of methylamine

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

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

Thermophysical properties and microstructure of graphite flake/copper composites processed by electroless copper coating

Energy Technology Data Exchange (ETDEWEB)

Liu, Qian; He, Xin-Bo; Ren, Shu-Bin; Zhang, Chen; Ting-Ting, Liu; Qu, Xuan-Hui, E-mail: quxh@ustb.edu.cn

2014-02-25

Highlights: • GF–copper composites were fabricated using a sparking plasma sintering, which involves coating GF with copper, using electroless plating technique. • The oriented graphite flake distributed homogeneously in matrix. • With the increase of flake graphite from 44 to 71 vol.%, the basal plane thermal conductivity of composites increases from 445 to 565 W m{sup −1} K{sup −1} and the thermal expansion of composites decreases from 8.1 to 5.0. • The obtained composites are suitable for electronic packaging materials. -- Abstract: This study focuses on the fabrication of thermal management material for power electronics applications using graphite flake reinforced copper composites. The manufacturing route involved electroless plating of copper on the graphite flake and further spark plasma sintering of composite powders. The relative density of the composites with 44–71 vol.% flakes achieved up to 98%. Measured thermal conductivities and coefficients of thermal expansion of composites ranged from 455–565 W m{sup −1} K{sup −1} and 8 to 5 ppm K{sup −1}, respectively. Obtained graphite flake–copper composites exhibit excellent thermophysical properties to meet the heat dispersion and matching requirements of power electronic devices to the packaging materials.

Thermal-hydraulic instabilities in pressure tube graphite - moderated boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Tsiklauri, G.; Schmitt, B.

1995-09-01

Thermally induced two-phase instabilities in non-uniformly heated boiling channels in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

Thermal-hydraulic instabilities in pressure tube graphite-moderated boiling water reactors

International Nuclear Information System (INIS)

Tsiklauri, G.; Schmitt, B.

1995-09-01

Thermally induced two-phase instabilities in non-uniformly heated boiling charmers in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement

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.

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.

The effective neutron temperature in heated graphite sleeves

Energy Technology Data Exchange (ETDEWEB)

Shaw, J A; Small, V G [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1963-08-15

In a series of oscillator measurements carried out in the reactor NERO the variation of the relative reaction rates of cadmium and boron absorbers has been used to determine the effective neutron temperature inside heated graphite sleeves. This work extends the scope of similar oscillator measurements previously carried out in DIMPLE, in that the bulk moderator is now graphite as opposed to D{sub 2}O in the former case. (author)

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)

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

Characterization of radiation damage induced by swift heavy ions in graphite

Energy Technology Data Exchange (ETDEWEB)

Hubert, Christian

2016-05-15

Graphite is a classical material in neutron radiation environments, being widely used in nuclear reactors and power plants as a moderator. For high energy particle accelerators, graphite provides ideal material properties because of the low Z of carbon and its corresponding low stopping power, thus when ion projectiles interact with graphite is the energy deposition rather low. This work aims to improve the understanding of how the irradiation with swift heavy ions (SHI) of kinetic energies in the range of MeV to GeV affects the structure of graphite and other carbon-based materials. Special focus of this project is given to beam induced changes of thermo-mechanical properties. For this purpose the Highly oriented pyrolytic graphite (HOPG) and glassy carbon (GC) (both serving as model materials), isotropic high density polycrystalline graphite (PG) and other carbon based materials like carbon fiber carbon composites (CFC), chemically expanded graphite (FG) and molybdenum carbide enhanced graphite composites (MoC) were exposed to different ions ranging from {sup 131}Xe to {sup 238}U provided by the UNILAC accelerator at GSI in Darmstadt, Germany. To investigate structural changes, various in-situ and off-line measurements were performed including Raman spectroscopy, x-ray diffraction and x-ray photo-electron spectroscopy. Thermo-mechanical properties were investigated using the laser-flash-analysis method, differential scanning calorimetry, micro/nano-indentation and 4-point electrical resistivity measurements. Beam induced stresses were investigated using profilometry. Obtained results provided clear evidence that ion beam-induced radiation damage leads to structural changes and degradation of thermal, mechanical and electrical properties of graphite. PG transforms towards a disordered sp2 structure, comparable to GC at high fluences. Irradiation-induced embrittlement is strongly reducing the lifetime of most high-dose exposed accelerator components. For

The relationship between irradiation induced dimensional change and the coefficient of thermal expansion: A new look

International Nuclear Information System (INIS)

Hall, G.; Marsden, B.J.; Fok, A.; Smart, J.

2002-01-01

In the 1960s, J.H.W. Simmons derived a theoretical relationship between the coefficient of thermal expansion (CTE) and dimensional changes in irradiated graphite. At low irradiation dose, the theory was shown to be consistent with experimental observations. However, at higher doses the results diverge. Despite this, modified versions of this theory have been used as the basis of the design and life prediction calculations for graphite-moderated reactors. This paper revisits Simmons's theory, summarising the assumptions made in its derivation. The paper then modifies and applies the theory to the dimensional change and CTE change behaviour in isotropic nuclear graphite, making use of trends in irradiated behaviour recently derived using finite element analyses. The importance of these issues to present HTR technology is that the life of HTR graphite components is related to their irradiated dimensional change behaviour. A more in depth understanding of this behaviour will allow suitable graphite material to be selected or new graphite types to be developed. (author)

Graphite limiter and armour damage in Doublet III

International Nuclear Information System (INIS)

McKelvey, T.; Taylor, T.; Trester, P.

1983-01-01

Graphite coated with TiC has been used extensively in Doublet III for limiters and neutral beam armour. Performance of these components has been superior to that of the metal components previously used. Damage to the coated graphite has occurred and can be classified into three categories: (1) gross failure of the graphite due to thermal stresses induced by the combination of high applied energy fluxes and mechanical restraint, (2) surface failure of the graphite due to runaway electron impingement, and (3) loss of TiC coating due to arcing, sputtering, vaporization and spalling, primarily during plasma disruptions and other abnormal plasma conditions. Design improvements are being continually implemented to minimize this damage and its consequences. (author)

Improvement in transdermal drug delivery performance by graphite oxide/temperature-responsive hydrogel composites with micro heater

Energy Technology Data Exchange (ETDEWEB)

Yun, Jumi [Department of Fine Chemical Engineering and Applied Chemistry, BK21-E2M, Chungnam National University, Daejeon 305-764 (Korea, Republic of); Lee, Dae Hoon [Environment Research Division, Korea Institute of Machinery and Materials, 171 Jang-dong, Yusong-gu, Daejeon 305-343 (Korea, Republic of); Im, Ji Sun [Department of Fine Chemical Engineering and Applied Chemistry, BK21-E2M, Chungnam National University, Daejeon 305-764 (Korea, Republic of); Kim, Hyung-Il, E-mail: hikim@cnu.ac.kr [Department of Fine Chemical Engineering and Applied Chemistry, BK21-E2M, Chungnam National University, Daejeon 305-764 (Korea, Republic of)

2012-08-01

Transdermal drug delivery system (TDDS) was prepared with temperature-responsive hydrogel. The graphite was oxidized and incorporated into hydrogel matrix to improve the thermal response of hydrogel. The micro heater was fabricated to control the temperature precisely by adopting a joule heating method. The drug in hydrogel was delivered through a hairless mouse skin by controlling temperature. The efficiency of drug delivery was improved obviously by incorporation of graphite oxide due to the excellent thermal conductivity and the increased interfacial affinity between graphite oxide and hydrogel matrix. The fabricated micro heater was effective in controlling the temperature over lower critical solution temperature of hydrogel precisely with a small voltage less than 1 V. The cell viability test on graphite oxide composite hydrogel showed enough safety for using as a transdermal drug delivery patch. The performance of TDDS could be improved noticeably based on temperature-responsive hydrogel, thermally conductive graphite oxide, and efficient micro heater. - Graphical abstract: The high-performance transdermal drug delivery system could be prepared by combining temperature-responsive hydrogel, thermally conductive graphite oxide with improved interfacial affinity, and efficient micro heater fabricated by a joule heating method. Highlights: Black-Right-Pointing-Pointer High performance of transdermal drug delivery system with an easy control of voltage. Black-Right-Pointing-Pointer Improved thermal response of hydrogel by graphite oxide incorporation. Black-Right-Pointing-Pointer Efficient micro heater fabricated by a joule heating method.

Changes in the physical and mechanical properties of graphite on irradiation in ditolylmethane

International Nuclear Information System (INIS)

Gavrilin, A.I.; Lebedev, I.G.; Sudakova, N.V.; Rizvanov, V.K.

1987-01-01

Results are presented from the irradiation and mechanical and structural testing of four grades of graphite - GMZ, VPG, MPG-6, and PG-50 - for use as moderator materials in organic cooled and graphite moderated reactors. Irradiation was carried out in the ARBUS-AST-1 reactor. Photomicrography was used to determine pore structure and ultimate strength in bending and compression was determined mechanically. Irradiation was found to increase the strength of GMZ, PMG-6, and PG-50 considerably, due to the healing of microdefects as a result of the pores filling with radiolysis products from the coolant, ditolylmethane. Conversely, VPG graphite, which has closed porosity, lost strength on irradiation

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)

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

Simultaneous heating and compression of irradiated graphite during synchrotron microtomographic imaging

Science.gov (United States)

Bodey, A. J.; Mileeva, Z.; Lowe, T.; Williamson-Brown, E.; Eastwood, D. S.; Simpson, C.; Titarenko, V.; Jones, A. N.; Rau, C.; Mummery, P. M.

2017-06-01

Nuclear graphite is used as a neutron moderator in fission power stations. To investigate the microstructural changes that occur during such use, it has been studied for the first time by X-ray microtomography with in situ heating and compression. This experiment was the first to involve simultaneous heating and mechanical loading of radioactive samples at Diamond Light Source, and represented the first study of radioactive materials at the Diamond-Manchester Imaging Branchline I13-2. Engineering methods and safety protocols were developed to ensure the safe containment of irradiated graphite as it was simultaneously compressed to 450N in a Deben 10kN Open-Frame Rig and heated to 300°C with dual focused infrared lamps. Central to safe containment was a double containment vessel which prevented escape of airborne particulates while enabling compression via a moveable ram and the transmission of infrared light to the sample. Temperature measurements were made in situ via thermocouple readout. During heating and compression, samples were simultaneously rotated and imaged with polychromatic X-rays. The resulting microtomograms are being studied via digital volume correlation to provide insights into how thermal expansion coefficients and microstructure are affected by irradiation history, load and heat. Such information will be key to improving the accuracy of graphite degradation models which inform safety margins at power stations.

Measurement and investigation of effects of coal tar pitch fractions in nuclear graphite properties

International Nuclear Information System (INIS)

Fatemi, K.; Fatoorehchian, S.; Ahari Hashemi, F.; Ahmadi, Sh.

2003-01-01

Coal tar pitch has a complex chemical structure. Determination of α, β, γ fractions, is one of the methods to get information about its properties. In graphite fabrication it plays a role as a binder for coke particles. During the thermal treatment it carbonizes and changes to a secondary coke. This has considerable affects on the graphite properties. In this paper, determination of α, β, γ-1 fraction in three different types of pitches have been carried out. Graphite specimens have been fabricated by using these pitches and anisotropy coke in laboratory scale. The graphite properties have been compared with the nuclear graphite prototype. The comparison of the results showed that the density and compression strength are appreciable while the anisotropy factor of properties is about one. The linear thermal expansion in graphite from Iranian pitch had a better, result, where it stands in the nuclear range of usage. As a result, our studies showed that the graphite properties are affected by properties of pitch fractions, where it can be used as a proper sample for the graphite fabrication

Thermal-hydraulic experiments and analyses on cold moderator

International Nuclear Information System (INIS)

Aso, Tomokazu; Kaminaga, Masanori; Terada, Atsushi; Hino, Ryutaro

2001-01-01

A cold moderator using supercritical hydrogen is one of the key components in a MW-scale spallation target system, which directly affects the neutronic performance both in intensity and resolution. Since a hydrogen temperature rise in the moderator vessel affects the neutronic performance, it is necessary to suppress the local temperature rise within 3 K. In order to develop the conceptual design of the moderator structure in progress, the flow patterns were measured using a PIV (Particle Image Velocimeter) system under water flow conditions using a flat model that simulated a moderator vessel. From these results, the flow patterns (such as recirculation flows, stagnant flows etc.) were clarified. The hydraulic analytical results obtained using the STAR-CD code agreed well with experimental results. Thermal-hydraulic analyses in the moderator vessel were carried out using the STAR-CD code. Based on these results, we clarified the possibility of suppressing the local temperature rise to within 3 K under 2 MW operating conditions. In order to achieve the cost decreasing of the hydrogen loop, it is necessary to operate it reducing the hydrogen flow rate and the whole hydrogen mass. Then improved moderator concept using blowholes and a twisted tape was proposed, and we have tried to examine the effect of the blowing flow from the inlet pipe. From the experimental and analytical results, the blowing flow could be feasible for the suppression of the stagnant region. (author)

New insights into canted spiro carbon interstitial in graphite

Science.gov (United States)

EL-Barbary, A. A.

2017-12-01

The self-interstitial carbon is the key to radiation damage in graphite moderator nuclear reactor, so an understanding of its behavior is essential for plant safety and maximized reactor lifetime. The density functional theory is applied on four different graphite unit cells, starting from of 64 carbon atoms up to 256 carbon atoms, using AIMPRO code to obtain the energetic, athermal and mechanical properties of carbon interstitial in graphite. This study presents first principles calculations of the energy of formation that prove its high barrier to athermal diffusion (1.1 eV) and the consequent large critical shear stress (39 eV-50 eV) necessary to shear graphite planes in its presence. Also, for the first time, the gamma surface of graphite in two dimensions is calculated and found to yield the critical shear stress for perfect graphite. Finally, in contrast to the extensive literature describing the interstitial of carbon in graphite as spiro interstitial, in this work the ground state of interstitial carbon is found to be canted spiro interstitial.

Thermal comfort, physiological responses and performance during exposure to a moderate temperature drift

DEFF Research Database (Denmark)

Schellen, Lisje; van Marken Lichtenbelt, Wouter; de Wit, Martin

2008-01-01

The objective of this research was to study the effects of a moderate temperature drift on human thermal comfort, physiological responses, productivity and performance. A dynamic thermophysiological model was used to examine the possibility of simulating human thermal responses and thermal comfort...... temperature corresponding with a neutral thermal sensation (control situation). During the experiments both physiological responses and thermal sensation were measured. Productivity and performance were assessed with a ‘Remote Performance Measurement’ (RPM) method. Physiological and thermal sensation data...

Scaled-Up Production and Transport Applications of Graphitic Carbon Nanomaterials

Science.gov (United States)

Saviers, Kimberly R.

Graphitic carbon nanomaterials enhance the performance of engineered systems for energy harvesting and storage. However, commercial availability remains largely cost-prohibitive due to technical barriers to mass production. This thesis examines both the scaled-up production and energy transport applications of graphitic materials. Cost driven-production of graphitic petals is developed, carbon nanotube array thermal interface materials enhance waste heat energy harvesting, and microsupercapacitors are visually examined using a new electroreflectance measurement method. Graphitic materials have previously been synthesized using batch-style processing methods with small sample sizes, limiting their commercial viability. In order to increase production throughput, a roll-to-roll radio-frequency plasma chemical vapor deposition method is employed to continuously deposit graphitic petals on carbon fiber tow. In consideration of a full production framework, efficient and informative characterization methods in the form of electrical resistance and electrochemical capacitance are highlighted. To co-optimize the functional characteristics of the material, the processing conditions are comprehensively varied using a data-driven predictive design of experiments method. Repeatable and reliable production of graphitic materials will enable a host of creative graphene-based devices to emerge into the marketplace. Two such applications are discussed in the remaining chapters. Waste heat is most efficiently harvested at high temperatures, such as vehicle exhaust systems near 600°C. However, the resistance to heat flux at the interfaces between the harvesting device and its surroundings is detrimental to the system-level performance. To study the performance of thermal interface materials up to 700°C, a reference bar measurement method was designed. Design considerations are discussed and compared to past implementations, particularly regarding radiation heat flux and thermal

Thermal and electrical conductivity enhancement of graphite nanoplatelets on form-stable polyethylene glycol/polymethyl methacrylate composite phase change materials

International Nuclear Information System (INIS)

Zhang, Lei; Zhu, Jiaoqun; Zhou, Weibing; Wang, Jun; Wang, Yan

2012-01-01

Graphite nanoplatelets (GnPs), obtained by sonicating the expanded graphite, were employed to simultaneously enhance the thermal (k) and electrical (σ) conductivity of organic form-stable phase change materials (FSPCMs). Using the method of in situ polymerization upon ultrasonic irradiation, GnPs serving as the conductive fillers and polyethylene glycol (PEG) acting as the phase change material (PCM) were uniformly dispersed and embedded inside the network structure of polymethyl methacrylate (PMMA), which contributed to the well package and self-supporting properties of composite FSPCMs. X-ray diffraction and Fourier transform infrared spectroscopy results indicated that the GnPs were physically combined with PEG/PMMA matrix and did not participate in the polymerization. The GnPs additives were able to effectively enhance the k and σ of organic FSPCM. When the mass ratio of GnP was 8%, the k and σ of FSPCM changed up to 9 times and 8 orders of magnitude over that of PEG/PMMA matrix, respectively. The improvements in both k and σ were mainly attributed to the well dispersion and large aspect ratio of GnPs, which were endowed with benefit of forming conducting network in polymer matrix. It was also confirmed that all the prepared specimens possessed available thermal storage density and thermal stability. -- Highlights: ► GnPs were employed to simultaneously enhance the k and σ of organic FSPCMs. ► PEG/PMMA/GnPs composite FSPCMs were prepared by in situ polymerization method. ► The composite FSPCMs exhibited well package and self-supporting properties. ► GnPs additives effectively enhanced the k and σ of composite FSPCMs. ► All the composites possessed available thermal storage density and thermal stability.

2-D CFD time-dependent thermal-hydraulic simulations of CANDU-6 moderator flows

Energy Technology Data Exchange (ETDEWEB)

Mehdi Zadeh, Foad [Department of Engineering Physics/Polytechnique Montréal, Montréal, QC (Canada); Étienne, Stéphane [Department of Mechanical Engineering/Polytechnique Montréal, Montréal, QC (Canada); Teyssedou, Alberto, E-mail: alberto.teyssedou@polymtl.ca [Department of Engineering Physics/Polytechnique Montréal, Montréal, QC (Canada)

2016-12-01

Highlights: • 2-D time-dependent CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • Frequency components indicate moderator flow oscillations vs. Richardson numbers. - Abstract: The distribution of the fluid temperature and mass density of the moderator flow in CANDU-6 nuclear power reactors may affect the reactivity coefficient. For this reason, any possible moderator flow configuration and consequently the corresponding temperature distributions must be studied. In particular, the variations of the reactivity may result in major safety issues. For instance, excessive temperature excursions in the vicinity of the calandria tubes nearby local flow stagnation zones, may bring about partial boiling. Moreover, steady-state simulations have shown that for operating condition, intense buoyancy forces may be dominant, which can trigger a thermal stratification. Therefore, the numerical study of the time-dependent flow transition to such a condition, is of fundamental safety concern. Within this framework, this paper presents detailed time-dependent numerical simulations of CANDU-6 moderator flow for a wide range of flow conditions. To get a better insight of the thermal-hydraulic phenomena, the simulations were performed by covering long physical-time periods using an open-source code (Code-Saturne V3) developed by Électricité de France. The results show not only a region where the flow is characterized by coherent structures of flow fluctuations but also the existence of two limit cases where fluid oscillations disappear almost completely.

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

Science.gov (United States)

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

2009-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-04-30

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Graphite development for gas-cooled reactors in the USA

International Nuclear Information System (INIS)

Burchell, T.D.

1991-01-01

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

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.

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.)

Performance assessment of turbulence models for the prediction of moderator thermal flow inside CANDU calandria

International Nuclear Information System (INIS)

Lee, Gong Hee; Bang, Young Seok; Woo, Sweng Woong

2012-01-01

The moderator thermal flow in the CANDU calandria is generally complex and highly turbulent because of the interaction of the buoyancy force with the inlet jet inertia. In this study, the prediction performance of turbulence models for the accurate analysis of the moderator thermal flow are assessed by comparing the results calculated with various types of turbulence models in the commercial flow solver FLUENT with experimental data for the test vessel at Sheridan Park Engineering Laboratory (SPEL). Through this comparative study of turbulence models, it is concluded that turbulence models that include the source term to consider the effects of buoyancy on the turbulent flow should be used for the reliable prediction of the moderator thermal flow inside the CANDU calandria

Influence of Particle Size on Properties of Expanded Graphite

Directory of Open Access Journals (Sweden)

Kurajica, S

2010-02-01

Full Text Available Expanded graphite has been applied widely in thermal insulation, adsorption, vibration damping, gasketing, electromagnetic interference shielding etc. It is made by intercalation of natural flake graphite followed by thermal expansion. Intercalation is a process whereby an intercalant material is inserted between the graphene layers of a graphite crystal. Exfoliation, a huge unidirectional expansion of the starting intercalated flakes, occurs when the graphene layers are forced apart by the sudden decomposition and vaporization of the intercalated species by thermal shock. Along with production methodologies, such as the intercalation process and heat treatment, the raw material characteristics, especially particle size, strongly influence the properties of the final product.This report evaluates the influence of the particle size of the raw material on the intercalation and expansion processes and consequently the properties of the exfoliated graphite. Natural crystalline flake graphite with wide particle diameter distribution (between dp = 80 and 425 µm was divided into four size-range portions by sieving. Graphite was intercalated via perchloric acid, glacial acetic acid and potassium dichromate oxidation and intercalation procedure. 5.0 g of graphite, 7.0 g of perchloric acid, 4.0 g of glacial acetic acid and 2.0 g of potassium dichromate were placed in glass reactor. The mixture was stirred with n = 200 min–1 at temperature of 45 °C during 60 min. Then it was filtered and washed with distilled water until pH~6 and dried at 60 °C during 24 h. Expansion was accomplished by thermal shock at 1000 °C for 1 min. The prepared samples were characterized by means of exfoliation volume measurements, simultaneous differential thermal analysis and thermo-gravimetry (DTA/TGA, X-ray diffraction (XRD, Fourier transform infrared spectroscopy (FTIR, BET measurements and scanning electron microscopy (SEM.X-ray diffraction indicated a change of distance

Effect of electrical pulse treatment on the thermal fatigue resistance of bionic compacted graphite cast iron processed in water

International Nuclear Information System (INIS)

Liu, Yan; Zhou, Hong; Su, Hang; Yang, Chunyan; Cheng, Jingyan; Zhang, Peng; Ren, Luquan

2012-01-01

Highlights: ► Electrical pulse treatment can reduce cracks on bionic units before thermal fatigue tests. ► Electrical pulse treatment can reduce crack sources during thermal fatigue tests. ► Thermal fatigue resistance of bionic units processed in water is enhanced. ► Thermal fatigue resistance of bionic CGI processed in water is improved. -- Abstract: In order to further enhance the thermal fatigue resistance of bionic compacted graphite cast iron (CGI) which is processed by laser in water, the electrical pulse treatment is applied to improve the thermal fatigue resistance of bionic units. The results show that the electrical pulse treatment causes the supersaturated carbon atoms located in the lattice of austenite to react with the iron atoms to form the Fe 3 C. The microstructures of the bionic units processed in water are refined by the electrical pulse treatment. The cracks on the bionic units are reduced by the electrical pulse treatment before the thermal fatigue tests; and during the tests, the thermal fatigue resistance of bionic units is therefore enhanced by reducing the crack sources. By this way, the thermal fatigue resistance of bionic CGI processed in water is improved.

Thermomechanical behavior of graphite and coating materials subjected to a high heat flux

Energy Technology Data Exchange (ETDEWEB)

Ioki, K.; Yamada, M.; Nishikawa, M.; Uchikawa, T.; Onozuka, M.; Yamao, H.

1987-07-01

This study has been performed for the development of limiter and divertor plates. Their thermal and thermomechanical behavior were examined in heat load experiments with an electron beam facility, and were compared with analysis results. Graphite was proven to have a high thermal shock resistance. Its erosion thickness and thermal contact conductance were also studied. Copper alloy with coating and graphite brazed to metal were tested, and their feasibility was demonstrated for use as limiter and divertor plates of an advanced-type concept.

Chemical vapor deposition of TiB2 on graphite

International Nuclear Information System (INIS)

Pierson, H.O.; Randich, E.; Mattox, D.M.

1978-01-01

This study is an experimental investigation of the coating of graphite with TiB 2 by chemical vapor deposition (CVD) using the hydrogen reduction of BCl 3 and TiCl 4 at 925 0 C and 1 atm. Reasonable matching of the thermal expansion of TiB 2 and graphite was necessary to eliminate cracking. A suitable graphite was POCO DFP-1. Adhesion was improved by having a slightly rough graphite surface. Heat treatment at 2000 0 C and above resulted in a certain degree of diffusion. No melting or solid phases other than TiB 2 and graphite were detected up to 2400 0 C. The coatings showed no failure when repeatedly submitted to an electron beam pulse of 2 KW/cm 2 for 0.8 sec

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)

Preparation and Characterization of Graphite Waste/CeO2 Composites

Science.gov (United States)

Kusrini, E.; Utami, C. S.; Nasruddin; Prasetyanto, E. A.; Bawono, Aji A.

2018-03-01

In this research, the chemical modification of graphite waste with CeO2 was developed and characterized. Graphite waste was pretreated with mechanical to obtain the size 200 mesh (75 μm), and thermal methods at 110°C oven for 6 hours. Here, we demonstrate final properties of graphite before modification (GBM), activated graphite (GA) and graphite/CeO2 composite with variation of 0.5, 1 and 2 g of CeO2 (G0.5; G1; G2). The effect of CeO2 concentration was observed. The presence of cerium in modified graphite samples (G0.5; G1; G2) were analyzed using SEM-EDX. The results show that the best surface area was found in G2 is 26.82 m2/g. The presence of CeO2 onto graphite surface does not significantly increase the surface area of composites.

Recent developments in graphite. [Use in HTGR and aerospace

Energy Technology Data Exchange (ETDEWEB)

Cunningham, J.E.

1983-01-01

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

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)

An optimization study of peak thermal neutron flux in moderators of advanced repetitive pulse reactors

International Nuclear Information System (INIS)

Asaoka, Takumi; Watanabe, N.

1976-01-01

In achieving a high peak thermal neutron flux in hydrogenous moderators installed in repetitive pulse reactors, the core-moderator arrangement can play as much an important role as the moderator design itself. However, the effect of the former has not been adequately emphasized to date, while a rather extensive study has been made on the latter. The present study concerns with a core-moderator system parameter optimization for a repetitive accelerator pulsed fast reactor. The results have shown that small differences in the arrangement resulting from the optimizations of various parameters are significant and the effects can be summed up to give an increase in the peak thermal flux by a factor of about two. (auth.)

Thermomechanical behavior of graphite and coating materials subjected to a high heat flux

International Nuclear Information System (INIS)

Ioki, K.; Yamada, M.; Nishikawa, M.; Uchikawa, T.; Onozuka, M.; Yamao, H.

1987-01-01

This study has been performed for the development of limiter and divertor plates. Their thermal and thermomechanical behavior were examined in heat load experiments with an electron beam facility, and were compared with analysis results. Graphite was proven to have a high thermal shock resistance. Its erosion thickness and thermal contact conductance were also studied. Copper alloy with coating and graphite brazed to metal were tested, and their feasibility was demonstrated for use as limiter and divertor plates of an advanced-type concept. (orig.)

Repairing the deteriorated thermal insulation in the serpentine - moderator tank - SLCD assemblies

International Nuclear Information System (INIS)

Gyongyosi, Tiberiu

2004-01-01

Deterioration during operation of the thermal insulation at the upper serpentines in the serpentine assembly in the moderator tank of SLCD (the system of localising the failed fuel) can create problems when one scans the fuel channels in case of failure of one of the ventilated air refrigerator in the rooms of the LAC 10 reactor. Recovering the thermal insulation is absolutely necessary but it is difficult to execute because the loading operation with the granulated layer of diatomaceous filtering agent must be effected directly on the moderator tank after some 24 h from the reactor shut down. The paper presents two possible methods of repairing together with the necessary technological facilities

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

International Nuclear Information System (INIS)

Anfimov, S.S.

2000-01-01

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

Irradiation behavior of graphite shielding materials for FBR

International Nuclear Information System (INIS)

Maruyama, Tadashi; Kaito, Takeji; Onose, Shoji; Shibahara, Itaru

1994-01-01

Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor 'JOYO' to fluences from 2.11 to 2.86x10 26 n/m 2 (E>0.1 MeV) at temperatures from 549 to 597degC. Postirradiation examination was carried out on dimensional change, elastic modulus, and the thermal conductivity. The result of measurement of dimensional change indicated that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased to two to three times of unirradiated values. A large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependency on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, but the change in specific heat was negligibly small. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens. (author)

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.

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

Preparation and thermal energy storage properties of paraffin/expanded graphite composite phase change material

International Nuclear Information System (INIS)

Zhang, Zhengguo; Zhang, Ni; Peng, Jing; Fang, Xiaoming; Gao, Xuenong; Fang, Yutang

2012-01-01

Highlights: ► EG was obtained by microwave irradiation to prepare the paraffin/EG composite PCM. ► Composite PCM was characterized by XRD to investigate the chemical compatibility. ► Temperature profiles of the composite PCM were obtained during thermal energy storage. -- Abstract: The paraffin/expanded graphite (EG) composite phase change material (PCM) was prepared by absorbing liquid paraffin into EG, in which paraffin was chosen as the PCM. EG was produced by microwave irradiation performed at room temperature. It was found that the EG prepared at 800 W irradiation power for 10 s exhibited the maximum sorption capacity of 92 wt% for paraffin. Scanning electron microscopy images showed that paraffin was uniformly dispersed in the pores of EG. Differential scanning calorimeter analysis indicated that the melting temperature of the composite PCM was close to that of paraffin, and its latent heat was equivalent to the calculated value based on the mass fraction of paraffin in the composite. X-ray diffraction analysis showed that the composite PCM was just a combination of paraffin with EG, and no new substance was produced. Thermal energy storage performance of the composite PCM was tested in a latent thermal energy storage (LTES) system. Transients of axial and radial temperature profiles were obtained in the LTES for the composite PCM and paraffin. The thermal energy storage charging duration for the composite PCM was reduced obviously compared to paraffin.

Studies on the development of special graphite for use in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Bhatia, G.; Aggarwal, R.K.; Saha, M.; Sengupta, P.R.; Mishra, A. [National Physical Lab., New Delhi (India). Carbon Technology Unit

2002-07-01

Special graphite is considered as a critical component of the present-day tokamaks wherein it acts as the armour material for plasma-facing components. This graphite is required to possess, besides other characteristics, high values of bulk density, bending strength and electrical and thermal conductivities and a low value of ash content. Since such graphite was not commercially available in the country, efforts to develop it were initiated at the National Physical Laboratory, New Delhi. The basic approach to develop this graphite was based on green coke method of making the high density graphite, wherein the green coke was modified by incorporating in it small amounts of conducting carbon materials, i.e. needle coke, synthetic graphite and natural graphite. The resulting graphites were characterized with respect to various physical characteristics, namely, green density, weight loss, volume shrinkage, linear shrinkage, bulk density, bending strength, Young's modulus and electrical resistivity, etc. The results are described and discussed in the present paper. 6 refs., 2 tabs.

High heat flux experiment on isotropic graphite using pulsed laser beam

International Nuclear Information System (INIS)

Kizaki, Hiroshi; Tokunaga, Kazutoshi; Fukuda, Shigehisa; Yoshida, Naoaki; Muroga, Takeo.

1989-01-01

In order to examine the plasma-withstanding behavior of isotropic graphite which is the leading favorite material for the first wall of nuclear fusion reactors, the pulsed thermal loading experiment was carried out by using a laser. As the result of analyzing the gas which was emitted during the pulsed thermal loading, together with the formation and release of various hydrocarbon gases, also the formation of carbon clusters due to the sublimation of carbon was observed. The vacuum characteristics and the dependence on thermal loading condition and surface treatment condition of these released gases were determined, and the problems and the way of improvement in its application to nuclear fusion reactors were elucidated. Since the isotropic graphite is of low atomic number, the radiation loss in plasma is small, and the improvement of the plasma parameters can be expected. Besides, the heat resistance and high temperature stability in vacuum are good, and the induced radioactivity is low. On the other hand, the quantity of gas occlusion is much, various hydrocarbon gases are formed at high temperature, and the wear due to sublimation arises by very high thermal loading. The experimental method, the observation of graphite surface by SEM, and the effect of carbon coating due to thermal decomposition are reported. (K.I.)

An experimental study of dynamic behaviour of graphite-polycarbonatediol polyurethane composites for protective coatings

Science.gov (United States)

Gómez, C. M.; Culebras, M.; Cantarero, A.; Redondo-Foj, B.; Ortiz-Serna, P.; Carsí, M.; Sanchis, M. J.

2013-06-01

Segmented polycarbonatediol polyurethane (PUPH) has been synthesized and modified with different amounts of graphite conductive filler (from 0 to 50 wt%). Thermal and dynamical thermal analysis of the composites clearly indicates changes in the polyurethane relaxations upon addition of graphite. Broadband dielectric spectroscopy has been used to study the dielectric properties of the (PUPH) and one composite in the frequency range from 10-2 to 107 Hz and in the temperature window of -140 to 170 °C. Relaxation processes associated with different molecular motions and conductivity phenomena (Maxwell-Wagner-Sillars and electrode polarization) are discussed and related to the graphite content.

Measurements of print-through in graphite fiber epoxy composites

Science.gov (United States)

Jaworske, Donald A.; Jeunnette, Timothy T.; Anzic, Judith M.

1989-01-01

High-reflectance accurate-contour mirrors are needed for solar dynamic space power systems. Graphite fiber epoxy composites are attractive candidates for such applications owing to their high modulus, near-zero coefficient of thermal expansion, and low mass. However, mirrors prepared from graphite fiber epoxy composite substrates often exhibit print-through, a distortion of the surface, which causes a loss in solar specular reflectance. Efforts to develop mirror substrates without print-through distortion require a means of quantifying print-through. Methods have been developed to quantify the degree of print-through in graphite fiber epoxy composite specimens using surface profilometry.

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.

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

International Nuclear Information System (INIS)

Knowles, A.N.

1980-01-01

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

The preliminary feasibility of intercalated graphite railgun armatures

International Nuclear Information System (INIS)

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

1991-01-01

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

Effects of graphite on rheological and conventional properties of bituminous binders

Directory of Open Access Journals (Sweden)

Yunus Erkuş

2017-07-01

Full Text Available In this study, the effects of graphite used for developing the rheological and conventional properties of bitumen were investigated using various bituminous binder tests. Penetration, softening point, rotational viscosity (RV, dynamic shear rheometer (DSR and bending beam rheometer (BBR tests were applied to bituminous binders modified with four different proportions of graphite by bitumen weight. The penetration values declined while softening point values increased with rising graphite content. While graphite induced 8 °C increases in mixing-compacting temperature by increasing the viscosity values, it also increased the rutting parameter. According to the BBR test, the deformation and stiffness values changed significantly with increasing graphite content, but the m-values did not change significantly. These results showed that graphite generally used for improving the thermal properties can improve to high temperature performance of mixtures. Keywords: Graphite, Bitumen, Conventional properties, Rheological properties

Measurements of anomalous neutron transport in bulk graphite

International Nuclear Information System (INIS)

Bowman, C.D.; Smith, G.A.; Vogelaar, B.; Howell, C.R.; Bilpuch, E.G.; Tornow, W.

2003-01-01

The neutron absorption of bulk granular graphite has been measured in a classical exponential diffusion experiment. Our first measurements of April 2002 implementing both exponential decay and pulsed die-away experiments and using the TUNL pulsed accelerator at Duke University as a neutron source indicated a capture cross section for graphite a striking factor of three lower than the measured value for carbon of 3.4 millibarns. Therefore a new exponential experiment with an improved geometry enabling greater accuracy has been performed giving an apparent cross section for carbon in the form of bulk granular graphite of less than 0.5 millibarns. This result confirms our first result and is also consistent with less than one part per million of boron in our graphite. The bulk density of the graphite is 1.02 compared with the actual particle density of 1.60 indicating a packing fraction of 0.64 or a void fraction of 0.36. We suspect that the apparent suppression of absorption in bulk graphite may be associated with the strong coherent diffraction of neutrons that dominates neutron transport in graphite. Coherent diffraction has never been taken into account in graphite reactor design and no neutron transport code including general use codes such as MCNP incorporate diffraction effects even though diffraction dominates many practical thermal neutron transport problems. (orig.)

Measurements of anomalous neutron transport in bulk graphite

Energy Technology Data Exchange (ETDEWEB)

Bowman, C.D.; Smith, G.A. [ADNA Corp., Los Alamos, NM (United States); Vogelaar, B. [Virginia Tech., Blacksburg, VA (United States); Howell, C.R.; Bilpuch, E.G.; Tornow, W. [Triangle Univ. Nuclear Lab., Duke Univ., Durham, NC (United States)

2003-07-01

The neutron absorption of bulk granular graphite has been measured in a classical exponential diffusion experiment. Our first measurements of April 2002 implementing both exponential decay and pulsed die-away experiments and using the TUNL pulsed accelerator at Duke University as a neutron source indicated a capture cross section for graphite a striking factor of three lower than the measured value for carbon of 3.4 millibarns. Therefore a new exponential experiment with an improved geometry enabling greater accuracy has been performed giving an apparent cross section for carbon in the form of bulk granular graphite of less than 0.5 millibarns. This result confirms our first result and is also consistent with less than one part per million of boron in our graphite. The bulk density of the graphite is 1.02 compared with the actual particle density of 1.60 indicating a packing fraction of 0.64 or a void fraction of 0.36. We suspect that the apparent suppression of absorption in bulk graphite may be associated with the strong coherent diffraction of neutrons that dominates neutron transport in graphite. Coherent diffraction has never been taken into account in graphite reactor design and no neutron transport code including general use codes such as MCNP incorporate diffraction effects even though diffraction dominates many practical thermal neutron transport problems. (orig.)

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

Colloidal graphite/graphene nanostructures using collagen showing enhanced thermal conductivity

Science.gov (United States)

Bhattacharya, Soumya; Dhar, Purbarun; Das, Sarit K; Ganguly, Ranjan; Webster, Thomas J; Nayar, Suprabha

2014-01-01

In the present study, the exfoliation of natural graphite (GR) directly to colloidal GR/graphene (G) nanostructures using collagen (CL) was studied as a safe and scalable process, akin to numerous natural processes and hence can be termed “biomimetic”. Although the exfoliation and functionalization takes place in just 1 day, it takes about 7 days for the nano GR/G flakes to stabilize. The predominantly aromatic residues of the triple helical CL forms its own special micro and nanoarchitecture in acetic acid dispersions. This, with the help of hydrophobic and electrostatic forces, interacts with GR and breaks it down to nanostructures, forming a stable colloidal dispersion. Surface enhanced Raman spectroscopy, X-ray diffraction, photoluminescence, fluorescence, and X-ray photoelectron spectroscopy of the colloid show the interaction between GR and CL on day 1 and 7. Differential interference contrast images in the liquid state clearly reveal how the GR flakes are entrapped in the CL fibrils, with a corresponding fluorescence image showing the intercalation of CL within GR. Atomic force microscopy of graphene-collagen coated on glass substrates shows an average flake size of 350 nm, and the hexagonal diffraction pattern and thickness contours of the G flakes from transmission electron microscopy confirm ≤ five layers of G. Thermal conductivity of the colloid shows an approximate 17% enhancement for a volume fraction of less than approximately 0.00005 of G. Thus, through the use of CL, this new material and process may improve the use of G in terms of biocompatibility for numerous medical applications that currently employ G, such as internally controlled drug-delivery assisted thermal ablation of carcinoma cells. PMID:24648728

Colloidal graphite/graphene nanostructures using collagen showing enhanced thermal conductivity.

Science.gov (United States)

Bhattacharya, Soumya; Dhar, Purbarun; Das, Sarit K; Ganguly, Ranjan; Webster, Thomas J; Nayar, Suprabha

2014-01-01

In the present study, the exfoliation of natural graphite (GR) directly to colloidal GR/graphene (G) nanostructures using collagen (CL) was studied as a safe and scalable process, akin to numerous natural processes and hence can be termed "biomimetic". Although the exfoliation and functionalization takes place in just 1 day, it takes about 7 days for the nano GR/G flakes to stabilize. The predominantly aromatic residues of the triple helical CL forms its own special micro and nanoarchitecture in acetic acid dispersions. This, with the help of hydrophobic and electrostatic forces, interacts with GR and breaks it down to nanostructures, forming a stable colloidal dispersion. Surface enhanced Raman spectroscopy, X-ray diffraction, photoluminescence, fluorescence, and X-ray photoelectron spectroscopy of the colloid show the interaction between GR and CL on day 1 and 7. Differential interference contrast images in the liquid state clearly reveal how the GR flakes are entrapped in the CL fibrils, with a corresponding fluorescence image showing the intercalation of CL within GR. Atomic force microscopy of graphene-collagen coated on glass substrates shows an average flake size of 350 nm, and the hexagonal diffraction pattern and thickness contours of the G flakes from transmission electron microscopy confirm ≤ five layers of G. Thermal conductivity of the colloid shows an approximate 17% enhancement for a volume fraction of less than approximately 0.00005 of G. Thus, through the use of CL, this new material and process may improve the use of G in terms of biocompatibility for numerous medical applications that currently employ G, such as internally controlled drug-delivery assisted thermal ablation of carcinoma cells.

Determining Thermal Specifications for Vegetated GREEN Roofs in Moderate Winter Climats

NARCIS (Netherlands)

Dr. Christoph Maria Ravesloot

2015-01-01

Because local weather conditions in moderate climates are changing constantly, heat transfer specifications of substrate and vegetation in vegetated green roofs also change accordingly. Nevertheless, it is assumed that vegetated green roofs can have a positive effect on the thermal performance of

A experimental system for the checking of the absorption of E.C.A.G. graphite; Empilement pour le controle du graphite E.C.A.G

Energy Technology Data Exchange (ETDEWEB)

Raievski, V; Vidal, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1958-07-01

A system is described for measuring the mean absorption cross section in thermal neutrons of graphite. This system consists of a graphite stack containing a Ra-Be source and a BF3 counter. A cavity in the stack receives the graphite to be studied or the graphite standard. By comparing the counting rates their absorption ratio can be deduced. The measurement is performed on graphite rods which have been machined before being placed in the pile. It provides the possibility of detecting over a batch of 1 ton of graphite, in a single measurement, a difference in absorption of 0.1 milli barn. (author) [French] On decrit un dispositif permettant de mesurer la section efficace moyenne d'absorption en neutrons thermiques du graphite. Ce dispositif est constitue par un empilement de graphite contenant une source de Ra-Be et un compteur a BF3. Une cavite menagee dans l'empilement peut recevoir le graphite a etudier ou le graphite etalon. Par comparaison des taux de comptage, on en deduit leur rapport d'absorption. La mesure est effectuee sur des barres de graphite usinees avant leur mise en place dans la pile. Elle permet de deceler sur un lot de graphite de 1 tonne, en une seule mesure, une difference d'absorption de 0,1 millibarn. (auteur)

Graphitic Carbon Foam Structural Cores and Multifunctional Applications

Data.gov (United States)

National Aeronautics and Space Administration — Graphitic carbon foams include a family of material forms and products with mechanical, thermal, and electrical properties that are tailor-able over a wide range....

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

Experiment and simulation of a LiFePO4 battery pack with a passive thermal management system using composite phase change material and graphite sheets

Science.gov (United States)

Lin, Chunjing; Xu, Sichuan; Chang, Guofeng; Liu, Jinling

2015-02-01

A passive thermal management system (TMS) for LiFePO4 battery modules using phase change material (PCM) as the heat dissipation source to control battery temperature rise is developed. Expanded graphite matrix and graphite sheets are applied to compensate low thermal conductivity of PCM and improve temperature uniformity of the batteries. Constant current discharge and mixed charge-discharge duties were applied on battery modules with and without PCM on a battery thermal characteristics test platform. Experimental results show that PCM cooling significantly reduces the battery temperature rise during short-time intense use. It is also found that temperature uniformity across the module deteriorates with the increasing of both discharge time and current rates. The maximum temperature differences at the end of 1C and 2C-rate discharges are both less than 5 °C, indicating a good performance in battery thermal uniformity of the passive TMS. Experiments on warm-keeping performance show that the passive TMS can effectively keep the battery within its optimum operating temperature for a long time during cold weather uses. A three dimensional numerical model of the battery pack with the passive TMS was conducted using ANSYS Fluent. Temperature profiles with respect to discharging time reveal that simulation shows good agreement with experiment at 1C-discharge rate.

Graphite curtain vacuum outgassing and heat transfer. Final report

International Nuclear Information System (INIS)

Fivel, H.J.; Lang, G.P.; Kipp, H.W.

1976-12-01

Thermal conductivity of a bundle of high conductivity graphite fibers (T-50) was measured as a function of temperature, density and fiber orientation at pressures of 10 -4 to 10 -5 torr. All 3 variables had a significant influence on thermal conductivity. The highest conductivity fiber bundle tested had a conductivity significantly less than dense, bulk nuclear grade graphite. The incorporation of heat pipes into a graphite spectral shaper will permit a 2-fold thicker shaper. Heat pipes not only increase the transport of heat within the spectral shaper but can increase heat transfer at the shaper-first wall interface and potentially serve as a means of attaching shaper modules to the first wall. A heat pipe using a liquid metal working fluid was fabricated and tested in magnetic fields of 1 and 2 Tesla. Liquid metal heat pipes can be used in a magnetic field of at least up to 2 Tesla. Much more work needs to be done to establish the capabilities for high performance heat pipes when used in magnetic fields. Four different types of graphite fibers were exposed in EBR-II to a neutron fluence of 3.5 x 10 21 cm -2 EFF at 470 0 C. Large axial shrinkages of 6.6 to 8.6% resulted

Experimental study on reactivity measurement in thermal reactor by polarity correlation method

International Nuclear Information System (INIS)

Yasuda, Hideshi

1977-11-01

Experimental study on the polarity correlation method for measuring the reactivity of a thermal reactor, especially the one possessing long prompt neutron lifetime such as graphite on heavy water moderated core, is reported. The techniques of reactor kinetics experiment are briefly reviewed, which are classified in two groups, one characterized by artificial disturbance to a reactor and the other by natural fluctuation inherent in a reactor. The fluctuation phenomena of neutron count rate are explained using F. de Hoffman's stochastic method, and correlation functions for the neutron count rate fluctuation are shown. The experimental results by polarity correlation method applied to the β/l measurements in both graphite-moderated SHE core and light water-moderated JMTRC and JRR-4 cores, and also to the measurement of SHE shut down reactivity margin are presented. The measured values were in good agreement with those by a pulsed neutron method in the reactivity range from critical to -12 dollars. The conditional polarity correlation experiments in SHE at -20 cent and -100 cent are demonstrated. The prompt neutron decay constants agreed with those obtained by the polarity correlation experiments. The results of experiments measuring large negative reactivity of -52 dollars of SHE by pulsed neutron, rod drop and source multiplication methods are given. Also it is concluded that the polarity and conditional polarity correlation methods are sufficiently applicable to noise analysis of a low power thermal reactor with long prompt neutron lifetime. (Nakai, Y.)

The Electrical and Thermal Conductivity of Woven Pristine and Intercalated Graphite Fiber-Polymer Composites

Science.gov (United States)

Gaier, James R.; Vandenburg, Yvonne Yoder; Berkebile, Steven; Stueben, Heather; Balagadde, Frederick

2002-01-01

A series of woven fabric laminar composite plates and narrow strips were fabricated from a variety of pitch-based pristine and bromine intercalated graphite fibers in an attempt to determine the influence of the weave on the electrical and thermal conduction. It was found generally that these materials can be treated as if they are homogeneous plates. The rule of mixtures describes the resistivity of the composite fairly well if it is realized that only the component of the fibers normal to the equipotential surface will conduct current. When the composite is narrow with respect to the fiber weave, however, there is a marked angular dependence of the resistance which was well modeled by assuming that the current follows only along the fibers (and not across them in a transverse direction), and that the contact resistance among the fibers in the composite is negligible. The thermal conductivity of composites made from less conductive fibers more closely followed the rule of mixtures than that of the high conductivity fibers, though this is thought to be an artifact of the measurement technique. Electrical and thermal anisotropy could be induced in a particular region of the structure by weaving together high and low conductivity fibers in different directions, though this must be done throughout all of the layers of the structure as interlaminar conduction precludes having only the top layer carry the anisotropy. The anisotropy in the thermal conductivity is considerably less than either that predicted by the rule of mixtures or the electrical resistivity.

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

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

Evaluation of hot spot factors for thermal and hydraulic design of HTTR

International Nuclear Information System (INIS)

Maruyama, So; Yamashita, Kiyonobu; Fujimoto, Nozomu; Murata, Isao; Sudo, Yukio; Murakami, Tomoyuki; Fujii, Sadao.

1993-01-01

High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal power and 950degC in reactor outlet coolant temperature. One of the major items in thermal and hydraulic design of the HTTR is to evaluate the maximum fuel temperature with a sufficient margin from a viewpoint of integrity of coated fuel particles. Hot spot factors are considered in the thermal and hydraulic design to evaluate the fuel temperature not only under the normal operation condition but also under any transient condition conservatively. This report summarizes the items of hot spot factors selected in the thermal and hydraulic design and their estimated values, and also presents evaluation results of the thermal and hydraulic characteristics of the HTTR briefly. (author)

Advanced Surface and Microstructural Characterization of Natural Graphite Anodes for Lithium Ion Batteries

Energy Technology Data Exchange (ETDEWEB)

Gallego, Nidia C [ORNL; Contescu, Cristian I [ORNL; Meyer III, Harry M [ORNL; Howe, Jane Y [ORNL; Meisner, Roberta Ann [ORNL; Payzant, E Andrew [ORNL; Lance, Michael J [ORNL; Yoon, Steve [A123 Systems, Inc.; Denlinger, Matthew [A123 Systems, Inc.; Wood III, David L [ORNL

2014-01-01

Natural graphite powders were subjected to a series of thermal treatments in order to improve the anode irreversible capacity loss (ICL) and capacity retention during long-term cycling of lithium ion batteries. A baseline thermal treatment in inert Ar or N2 atmosphere was compared to cases with a proprietary additive to the furnace gas environment. This additive substantially altered the surface chemistry of the natural graphite powders and resulted in significantly improved long-term cycling performance of the lithium ion batteries over the commercial natural graphite baseline. Different heat-treatment temperatures were investigated ranging from 950-2900 C with the intent of achieving the desired long-term cycling performance with as low of a maximum temperature and thermal budget as possible. A detailed summary of the characterization data is also presented, which includes X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), Raman spectroscopy, and temperature-programed desorption mass spectroscopy (TPD-MS). This characterization data was correlated to the observed capacity fade improvements over the course of long-term cycling at high charge-discharge rates in full lithium-ion coin cells. It is believed that the long-term performance improvements are a result of forming a more stable solid electrolyte interface (SEI) layer on the anode graphite surfaces, which is directly related to the surface chemistry modifications imparted by the proprietary gas environment during thermal treatment.

Synthesis of metal-fluoride nanoparticles supported on thermally reduced graphite oxide

Directory of Open Access Journals (Sweden)

Alexa Schmitz

2017-11-01

Full Text Available Metal-fluoride nanoparticles, (MFx-NPs with M = Fe, Co, Pr, Eu, supported on different types of thermally reduced graphite oxide (TRGO were obtained by microwave-assisted thermal decomposition of transition-metal amidinates, (M{MeC[N(iPr]2}n or [M(AMDn] with M = Fe(II, Co(II, Pr(III, and tris(2,2,6,6-tetramethyl-3,5-heptanedionatoeuropium, Eu(dpm3, in the presence of TRGO in the ionic liquid (IL 1-butyl-3-methylimidazolium tetrafluoroborate ([BMIm][BF4]. The crystalline phases of the metal fluorides synthesized in [BMIm][BF4] were identified by powder X-ray diffraction (PXRD to be MF2 for M = Fe, Co and MF3 for M = Eu, Pr. The diameters and size distributions of MFx@TRGO were from (6 ± 2 to (102 ± 41 nm. Energy-dispersive X-ray spectroscopy (EDX and X-ray photoelectron spectroscopy (XPS were used for further characterization of the MFx-NPs. Electrochemical investigations of the FeF2-NPs@TRGO as cathode material for lithium-ion batteries were evaluated by galvanostatic charge/discharge profiles. The results indicate that the FeF2-NPs@TRGO as cathode material can present a specific capacity of 500 mAh/g at a current density of 50 mA/g, including a significant interfacial charge storage contribution. The obtained nanomaterials show a good rate capacity as well (220 mAh/g and 130 mAh/g at a current density of 200 and 500 mA/g, respectively.

Physics of Plutonium Recycling in Thermal Reactors

International Nuclear Information System (INIS)

Kinchin, G.H.

1967-01-01

A substantial programme of experimental reactor physics work with plutonium fuels has been carried out in the UK; the purpose of this paper is to review the experimental and theoretical work, with emphasis on plutonium recycling in thermal reactors. Although the main incentive for some of the work may have been to study plutonium build-up in uranium-fuelled reactors, it is nevertheless relevant to plutonium recycling and no distinction is drawn between build-up and enrichment studies. A variety of techniques have been for determining reactivity, neutron spectrum and reaction rates in simple assemblies of plutonium-aluminium fuel with water, graphite and beryllia moderators. These experiments give confidence in the basic data and methods of calculation for near-homogeneous mixtures of plutonium and moderator. In the practical case of plutonium recycling it is necessary to confirm that satisfactory predictions can be made for heterogeneous lattices enriched with plutonium. In this field, experiments have been carried out with plutonium-uranium metal and oxide-cluster fuels in graphite-moderated lattices and in SGHW lattices, and the effects of 240 Pu have been studied by perturbation measurements with single fuel elements. The exponential and critical experiments have used tonne quantities of fuel with plutonium contents ranging from 0.25 to 1.2% and the perturbation experiments have extended both the range of plutonium contents and the range of isotopic compositions of plutonium. In addition to reactivity and reactivity coefficients, such as the temperature coefficients, attention has been concentrated on relative reaction rate distributions which provide evidence for variations of neutron spectrum. .Theoretical comparisons, together with similar comparisons for non-uniform lattices, establish the validity of methods of calculation which have been used to study the feasibility of plutonium recycling in thermal reactors. (author)

Physics of Plutonium Recycling in Thermal Reactors

Energy Technology Data Exchange (ETDEWEB)

Kinchin, G. H. [Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1967-09-15

A substantial programme of experimental reactor physics work with plutonium fuels has been carried out in the UK; the purpose of this paper is to review the experimental and theoretical work, with emphasis on plutonium recycling in thermal reactors. Although the main incentive for some of the work may have been to study plutonium build-up in uranium-fuelled reactors, it is nevertheless relevant to plutonium recycling and no distinction is drawn between build-up and enrichment studies. A variety of techniques have been for determining reactivity, neutron spectrum and reaction rates in simple assemblies of plutonium-aluminium fuel with water, graphite and beryllia moderators. These experiments give confidence in the basic data and methods of calculation for near-homogeneous mixtures of plutonium and moderator. In the practical case of plutonium recycling it is necessary to confirm that satisfactory predictions can be made for heterogeneous lattices enriched with plutonium. In this field, experiments have been carried out with plutonium-uranium metal and oxide-cluster fuels in graphite-moderated lattices and in SGHW lattices, and the effects of {sup 240}Pu have been studied by perturbation measurements with single fuel elements. The exponential and critical experiments have used tonne quantities of fuel with plutonium contents ranging from 0.25 to 1.2% and the perturbation experiments have extended both the range of plutonium contents and the range of isotopic compositions of plutonium. In addition to reactivity and reactivity coefficients, such as the temperature coefficients, attention has been concentrated on relative reaction rate distributions which provide evidence for variations of neutron spectrum. .Theoretical comparisons, together with similar comparisons for non-uniform lattices, establish the validity of methods of calculation which have been used to study the feasibility of plutonium recycling in thermal reactors. (author)

Experiences in the emptying of waste silos containing solid nuclear waste from graphite- moderated reactors

International Nuclear Information System (INIS)

Wall, S.; Schwarz, T.

2003-01-01

Before reactor sites can be handed over for ultimate decommissioning, at some sites silos containing waste from operations need to be emptied. The form and physical condition of the waste demands sophisticated retrieval technologies taking into account the onsite situation in terms of infrastructure and silo geometry. Furthermore, in the case of graphite moderated reactors, this waste usually includes several tonnes of graphite waste requiring special HVAC and dust handling measures. RWE NUKEM Group has already performed several contracts dealing with such emptying tasks. Of particular interest for the upcoming decommissioning projects in France might be the activities at Vandellos, Spain and Trawsfynnyd, UK. Retrieval System for Vandellos NPP is discussed. Following an international competitive tender exercise, RWE NUKEM won the contract to provide a turn-key retrieval system. This involved the design, manufacture and installation of a system built around the modules of a 200 kg capacity version of the ARTISAN manipulator system. The ARTISAN 200 manipulator, with remote slave arm detach facility, was deployed on a telescopic mast inserted into the silos through the roof penetrations. The manipulator deployed a range of tools to gather the waste and load it into a transfer basket, deployed through an adjacent penetration. After commissioning, the system cleared the vaults in less than the scheduled period with no failures. At the Trawsfynnyd Magnox plants two types of intermediate level waste (ILW) accumulated on site; namely Miscellaneous Activated Components (MAC) and Fuel Element Debris (FED). MAC is predominantly components that have been activated by the reactor core and then discharged. FED mainly consists of fuel cladding produced when fuel elements were prepared for dispatch to the reprocessing facility. RWE NUKEM Ltd. was awarded a contract to design, supply, commission and operate equipment to retrieve, pack and immobilize the two waste streams. Major

LiFSI vs. LiPF6 electrolytes in contact with lithiated graphite: Comparing thermal stabilities and identification of specific SEI-reinforcing additives

International Nuclear Information System (INIS)

Eshetu, Gebrekidan Gebresilassie; Grugeon, Sylvie; Gachot, Grégory; Mathiron, David; Armand, Michel; Laruelle, Stephane

2013-01-01

Lithium bis(fluorosulfonyl) imide (LiFSI) is regarded as an alternative to the classical LiPF 6 salt in today's LiFePO 4 /graphite-based Li-ion batteries electrolyte owing to its slightly higher conductivity and lower fluorine content. In an attempt to better evaluate the safety issues, here we report the comparative study of the LiFSI and LiPF 6 based electrolyte/lithiated graphite interface thermal behavior. DSC measurements with LiFSI-based electrolyte reveal a sharp exotherm with large heat release though at higher onset and peak temperatures compared to LiPF 6 -based electrolyte. With the help of GC/MS, 19 F NMR and ESI-HRMS analyses, we assume that this highly energetic peak around 200 °C, which is dependant upon the lithium content, is mainly related to electrochemical reduction of FSI − anion. In a strategy to limit the probability and damage of thermal runaway event, electrolyte additives such as vinylene carbonate (VC), fluoro ethylene carbonate (FEC), di-isocyanato hexane (DIH) and toluene di-isocyanate (TDI) have been investigated and shown to significantly lower the energy associated with the exothermic phenomenon

The single-collision thermalization approximation for application to cold neutron moderation problems

International Nuclear Information System (INIS)

Ritenour, R.L.

1989-01-01

The single collision thermalization (SCT) approximation models the thermalization process by assuming that neutrons attain a thermalized distribution with only a single collision within the moderating material, independent of the neutron's incident energy. The physical intuition on which this approximation is based is that the salient properties of neutron thermalization are accounted for in the first collision, and the effects of subsequent collisions tend to average out statistically. The independence of the neutron incident and outscattering energy leads to variable separability in the scattering kernel and, thus, significant simplification of the neutron thermalization problem. The approximation also addresses detailed balance and neutron conservation concerns. All of the tests performed on the SCT approximation yielded excellent results. The significance of the SCT approximation is that it greatly simplifies thermalization calculations for CNS design. Preliminary investigations with cases involving strong absorbers also indicates that this approximation may have broader applicability, as in the upgrading of the thermalization codes

Effect of Graphite Nanosheets on Properties of Poly(3-hydroxybutyrate-co-3-hydroxyvalerate

Directory of Open Access Journals (Sweden)

Larissa Stieven Montagna

2017-01-01

Full Text Available The influence of different contents, 0.25, 0.50, and 1.00 wt%, of graphite nanosheets (GNS on the properties of poly(3-hydroxybutyrate-co-3-hydroxyvalerate (PHBV nanocomposites obtained by solution casting method has been studied. GNS were prepared by three steps: intercalation (chemical exfoliation, expansion (thermal treatment, and the GNS obtainment (physical treatment by ultrasonic exfoliation. X-ray diffraction (XRD, Raman spectroscopy, and field emission gun-scanning electron microscopy (FE-SEM showed that the physical, chemical, and thermal treatments preserved the graphite sheets structure. XRD and Raman results also showed that GNS were dispersed in the PHBV matrix. The degree of crystallinity (Xc of the nanocomposites did not change when the graphite nanosheets were added. However, the GNS acted as nucleation agent for crystallization; that is, in the second heating the samples containing GNS showed two melting peaks. The addition the GNS did not change the thermal stability of the PHBV.

Graphite fiber/copper composites prepared by spontaneous infiltration

Science.gov (United States)

Wang, Hongbao; Tao, Zechao; Li, Xiangfen; Yan, Xi; Liu, Zhanjun; Guo, Quangui

2018-05-01

The major bottleneck in developing graphite fiber reinforced copper (GF/Cu) composites is the poor wettability of Cu/graphite system. Alloying element of chromium (Cr) is introduced to improve the wettability of liquid copper on graphite. Sessile drop method experiments illustrate that the contact angle of liquid Cu-Cr (1.0 wt.%) alloy on graphite substrate decreases to 43° at 1300 °C. The improvement of wettability is related to the formation of chromium carbide layer at interface zone. Based on the wetting experiment, a spontaneous infiltration method for preparing GF/Cu composites is proposed. Unidirectional GF preforms are infiltrated by Cu-Cr alloys without external pressure in a tubular furnace. Results reveal that the GF preform can be fully infiltrated by Cu-Cr alloy (8 wt.%) spontaneously when fiber volume fraction is 40%. The coefficient of thermal expansion (CTE) of GF/Cu-Cr (8.0 wt.%) composites is 4.68 × 10-6/K along the longitudinal direction.

Fabrication and testing of fire resistant graphite composite panels

Science.gov (United States)

Roper, W. D.

1986-01-01

Eight different graphite composite panels were fabricated using four different resin matrices. The resin matrices included Hercules 71775, a blend of vinylpolystyrpyridine and bismaleimide, H795, a bismaleimide, Cycom 6162, a phenolic, and PSP 6022m, a polystyrylpyridine. Graphite panels were fabricated using fabric or unidirectional tape. Described are the processes for preparing these panels and some of their mechanical, thermal and flammability properties. Panel properties are compared with state-of-the-art epoxy fiberglass composite panels.

Effect of airborne contaminants on the wettability of supported graphene and graphite.

Science.gov (United States)

Li, Zhiting; Wang, Yongjin; Kozbial, Andrew; Shenoy, Ganesh; Zhou, Feng; McGinley, Rebecca; Ireland, Patrick; Morganstein, Brittni; Kunkel, Alyssa; Surwade, Sumedh P; Li, Lei; Liu, Haitao

2013-10-01

It is generally accepted that supported graphene is hydrophobic and that its water contact angle is similar to that of graphite. Here, we show that the water contact angles of freshly prepared supported graphene and graphite surfaces increase when they are exposed to ambient air. By using infrared spectroscopy and X-ray photoelectron spectroscopy we demonstrate that airborne hydrocarbons adsorb on graphitic surfaces, and that a concurrent decrease in the water contact angle occurs when these contaminants are partially removed by both thermal annealing and controlled ultraviolet-O3 treatment. Our findings indicate that graphitic surfaces are more hydrophilic than previously believed, and suggest that previously reported data on the wettability of graphitic surfaces may have been affected by unintentional hydrocarbon contamination from ambient air.

Effect of airborne contaminants on the wettability of supported graphene and graphite

Science.gov (United States)

Li, Zhiting; Wang, Yongjin; Kozbial, Andrew; Shenoy, Ganesh; Zhou, Feng; McGinley, Rebecca; Ireland, Patrick; Morganstein, Brittni; Kunkel, Alyssa; Surwade, Sumedh P.; Li, Lei; Liu, Haitao

2013-10-01

It is generally accepted that supported graphene is hydrophobic and that its water contact angle is similar to that of graphite. Here, we show that the water contact angles of freshly prepared supported graphene and graphite surfaces increase when they are exposed to ambient air. By using infrared spectroscopy and X-ray photoelectron spectroscopy we demonstrate that airborne hydrocarbons adsorb on graphitic surfaces, and that a concurrent decrease in the water contact angle occurs when these contaminants are partially removed by both thermal annealing and controlled ultraviolet-O3 treatment. Our findings indicate that graphitic surfaces are more hydrophilic than previously believed, and suggest that previously reported data on the wettability of graphitic surfaces may have been affected by unintentional hydrocarbon contamination from ambient air.

Dispersion stability and thermophysical properties of environmentally friendly graphite oil–based nanofluids used in machining

Directory of Open Access Journals (Sweden)

Yu Su

2016-01-01

Full Text Available As environmentally friendly cutting fluids, vegetable-based oil and ester oil are being more and more widely used in metal cutting industry. However, their cooling and lubricating properties are required to be further improved in order to meet more cooling and lubricating challenges in high-efficiency machining. Nanofluids with enhanced heat carrying and lubricating capabilities seem to give a promising solution. In this article, graphite oil–based nanofluids with LB2000 vegetable-based oil and PriEco6000 unsaturated polyol ester as base fluids were prepared by ultrasonically assisted two-step method, and their dispersion stability and thermophysical properties such as viscosity and thermal conductivity were experimentally and theoretically investigated at different ultrasonication times. The results indicate that graphite-PriEco6000 nanofluid showed better dispersion stability, higher viscosity, and thermal conductivity than graphite-LB2000 nanofluid, which made it more suitable for application in high-efficiency machining as coolant and lubricant. The theoretical classical models showed good agreement with the thermal conductivity values of graphite oil–based nanofluids measured experimentally. However, the deviation between the experimental values of viscosity and the theoretical models was relatively big. New empirical correlations were proposed for predicting the viscosity of graphite oil–based nanofluids at various ultrasonication times.

Preparation of graphite intercalation compounds containing oligo and polyethers

Science.gov (United States)

Zhang, Hanyang; Lerner, Michael M.

2016-02-01

Layered host-polymer nanocomposites comprising polymeric guests between inorganic sheets have been prepared with many inorganic hosts, but there is limited evidence for the incorporation of polymeric guests into graphite. Here we report for the first time the preparation, and structural and compositional characterization of graphite intercalation compounds (GICs) containing polyether bilayers. The new GICs are obtained by either (1) reductive intercalation of graphite with an alkali metal in the presence of an oligo or polyether and an electrocatalyst, or (2) co-intercalate exchange of an amine for an oligo or polyether in a donor-type GIC. Structural characterization of products using powder X-ray diffraction, Raman spectroscopy, and thermal analyses supports the formation of well-ordered, first-stage GICs containing alkali metal cations and oligo or polyether bilayers between reduced graphene sheets.Layered host-polymer nanocomposites comprising polymeric guests between inorganic sheets have been prepared with many inorganic hosts, but there is limited evidence for the incorporation of polymeric guests into graphite. Here we report for the first time the preparation, and structural and compositional characterization of graphite intercalation compounds (GICs) containing polyether bilayers. The new GICs are obtained by either (1) reductive intercalation of graphite with an alkali metal in the presence of an oligo or polyether and an electrocatalyst, or (2) co-intercalate exchange of an amine for an oligo or polyether in a donor-type GIC. Structural characterization of products using powder X-ray diffraction, Raman spectroscopy, and thermal analyses supports the formation of well-ordered, first-stage GICs containing alkali metal cations and oligo or polyether bilayers between reduced graphene sheets. Electronic supplementary information (ESI) available: Domain size, additional Raman spectra info, compositional calculation, and packing fractions. See DOI: 10.1039/c5

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

Mechanistic elucidation of thermal runaway in potassium-ion batteries

Science.gov (United States)

Adams, Ryan A.; Varma, Arvind; Pol, Vilas G.

2018-01-01

For the first time, thermal runaway of charged graphite anodes for K-ion batteries is investigated, using differential scanning calorimetry (DSC) to probe the exothermic degradation reactions. Investigated parameters such as state of charge, cycle number, surface area, and binder demonstrate strong influences on the DSC profiles. Thermal runaway initiates at 100 °C owing to KxC8 - electrolyte reactions, but the K-ion graphite anode evolves significantly less heat as compared to the analogous Li-ion system (395 J g-1 vs. 1048 J g-1). The large volumetric expansion of graphite during potassiation cracks the SEI layer, enabling contact and reaction of KC8 - electrolyte, which diminishes with cycle number due to continuous SEI growth. High surface area graphite decreases the total heat generation, owing to thermal stability of the K-ion SEI layer. These findings illustrate the dynamic nature of K-ion thermal runaway and its many contrasts with the Li-ion graphite system, permitting possible engineering solutions for safer batteries.

A experimental system for the checking of the absorption of E.C.A.G. graphite

International Nuclear Information System (INIS)

Raievski, V.; Vidal, R.

1958-01-01

A system is described for measuring the mean absorption cross section in thermal neutrons of graphite. This system consists of a graphite stack containing a Ra-Be source and a BF3 counter. A cavity in the stack receives the graphite to be studied or the graphite standard. By comparing the counting rates their absorption ratio can be deduced. The measurement is performed on graphite rods which have been machined before being placed in the pile. It provides the possibility of detecting over a batch of 1 ton of graphite, in a single measurement, a difference in absorption of 0.1 milli barn. (author) [fr

Moderate Thermal Stress Causes Active and Immediate Expulsion of Photosynthetically Damaged Zooxanthellae (Symbiodinium from Corals.

Directory of Open Access Journals (Sweden)

Lisa Fujise

Full Text Available The foundation of coral reef biology is the symbiosis between corals and zooxanthellae (dinoflagellate genus Symbiodinium. Recently, coral bleaching, which often results in mass mortality of corals and the collapse of coral reef ecosystems, has become an important issue around the world as coral reefs decrease in number year after year. To understand the mechanisms underlying coral bleaching, we maintained two species of scleractinian corals (Acroporidae in aquaria under non-thermal stress (27°C and moderate thermal stress conditions (30°C, and we compared the numbers and conditions of the expelled Symbiodinium from these corals. Under non-thermal stress conditions corals actively expel a degraded form of Symbiodinium, which are thought to be digested by their host coral. This response was also observed at 30°C. However, while the expulsion rates of Symbiodinium cells remained constant, the proportion of degraded cells significantly increased at 30°C. This result indicates that corals more actively digest and expel damaged Symbiodinium under thermal stress conditions, likely as a mechanism for coping with environmental change. However, the increase in digested Symbiodinium expulsion under thermal stress may not fully keep up with accumulation of the damaged cells. There are more photosynthetically damaged Symbiodinium upon prolonged exposure to thermal stress, and corals release them without digestion to prevent their accumulation. This response may be an adaptive strategy to moderate stress to ensure survival, but the accumulation of damaged Symbiodinium, which causes subsequent coral deterioration, may occur when the response cannot cope with the magnitude or duration of environmental stress, and this might be a possible mechanism underlying coral bleaching during prolonged moderate thermal stress.

Moderate Thermal Stress Causes Active and Immediate Expulsion of Photosynthetically Damaged Zooxanthellae (Symbiodinium) from Corals.

Science.gov (United States)

Fujise, Lisa; Yamashita, Hiroshi; Suzuki, Go; Sasaki, Kengo; Liao, Lawrence M; Koike, Kazuhiko

2014-01-01

The foundation of coral reef biology is the symbiosis between corals and zooxanthellae (dinoflagellate genus Symbiodinium). Recently, coral bleaching, which often results in mass mortality of corals and the collapse of coral reef ecosystems, has become an important issue around the world as coral reefs decrease in number year after year. To understand the mechanisms underlying coral bleaching, we maintained two species of scleractinian corals (Acroporidae) in aquaria under non-thermal stress (27°C) and moderate thermal stress conditions (30°C), and we compared the numbers and conditions of the expelled Symbiodinium from these corals. Under non-thermal stress conditions corals actively expel a degraded form of Symbiodinium, which are thought to be digested by their host coral. This response was also observed at 30°C. However, while the expulsion rates of Symbiodinium cells remained constant, the proportion of degraded cells significantly increased at 30°C. This result indicates that corals more actively digest and expel damaged Symbiodinium under thermal stress conditions, likely as a mechanism for coping with environmental change. However, the increase in digested Symbiodinium expulsion under thermal stress may not fully keep up with accumulation of the damaged cells. There are more photosynthetically damaged Symbiodinium upon prolonged exposure to thermal stress, and corals release them without digestion to prevent their accumulation. This response may be an adaptive strategy to moderate stress to ensure survival, but the accumulation of damaged Symbiodinium, which causes subsequent coral deterioration, may occur when the response cannot cope with the magnitude or duration of environmental stress, and this might be a possible mechanism underlying coral bleaching during prolonged moderate thermal stress.

Monte Carlo design study of a moderated {sup 252}Cf source for in vivo neutron activation analysis of aluminium

Energy Technology Data Exchange (ETDEWEB)

Lewis, D.G.; Natto, S.S.A.; Evans, C.J. [Swansea In Vivo Analysis and Cancer Research Group, Department of Physics, University of Wales, Swansea (United Kingdom); Ryde, S.J.S. [Swansea In Vivo Analysis and Cancer Research Group, Department of Medical Physics and Clinical Engineering, Singleton Hospital, Swansea (United Kingdom)

1997-04-01

The Monte Carlo computer code MCNP has been used to design a moderated 2{sup 52}Cf neutron source for in vivo neutron activation analysis of aluminium (Al) in the bones of the hand. The clinical motivation is the need to monitor l body burden in subjects with renal dysfunction, at risk of Al toxicity. The design involves the source positioned on the central axis at one end of a cylindrical deuterium oxide moderator. The moderator is surrounded by a graphite reflector, with the hand inserted at the end of the moderator opposing the source. For a 1 mg {sup 252}Cf source, 15 cm long x 20 cm radius moderator and 20 cm thick reflector, the estimated minimum detection limit is .5 mg Al for a 20 min irradiation, with an equivalent dose of 16.5 mSv to the hand. Increasing the moderator length and/or introducing a fast neutron filter (for example silicon) further reduces interference from fast-neutron-induced reactions on phosphorus in bone, at the expense of decreased fluence of the thermal neutrons which activate Al. Increased source strengths may be necessary to compensate for this decreased thermal fluence, or allow measurements to be made within an acceptable time limit for the comfort of the patient. (author)

An Investigation of Reverse Flotation Separation of Sericite from Graphite by Using a Surfactant: MF

Directory of Open Access Journals (Sweden)

Yangshuai Qiu

2016-06-01

Full Text Available In this paper, a surfactant, atlox4862 (formaldehyde condensate of methyl naphthalene sulfonic sodium salt (MF, was introduced as a depressant for reverse flotation separation of sericite from graphite. Natural flake graphite has a strong hydrophobic property. After interacting with MF, the graphite became moderately hydrophilic. The flotation results showed that MF had a depressing ability for both sericite and graphite and that the flotation separation of sericite from graphite was attributed to the different declining levels of recovery between graphite and sericite with increased MF concentration. For a pulp pH of 8 and a MF concentration of 250 mg/L, the recovery rates of sericite and graphite were 89.7% and 11.3%, respectively. The results of the FTIR spectra and zeta potential measurements demonstrated that the interaction of MF with graphite and sericite is mainly through electrostatic attraction. MF was preferred to adsorb on the surface of graphite, decreasing its zeta potential and improving its hydrophilicity more than that of sericite.

Near-field thermal radiation between hyperbolic metamaterials: Graphite and carbon nanotubes

Energy Technology Data Exchange (ETDEWEB)

Liu, X. L.; Zhang, R. Z.; Zhang, Z. M., E-mail: zhuomin.zhang@me.gatech.edu [G. W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, Georgia 30332 (United States)

2013-11-18

The near-field radiative heat transfer for two hyperbolic metamaterials, namely, graphite and vertically aligned carbon nanotubes (CNTs), is investigated. Graphite is a naturally existing uniaxial medium, while CNT arrays can be modeled as an effective anisotropic medium. Different hyperbolic modes can be separately supported by these materials in certain infrared regions, resulting in a strong enhancement in near-field heat transfer. It is predicted that the heat flux between two CNT arrays can exceed that between SiC plates at any vacuum gap distance and is about 10 times higher with a 10 nm gap.

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

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.

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

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.

Preparation and characterization of aminated graphite oxide for CO2 capture

International Nuclear Information System (INIS)

Zhao Yunxia; Ding Huiling; Zhong Qin

2012-01-01

Adsorption with solid sorbents is one of the most promising options for postcombustion carbon dioxide (CO 2 ) capture. In this study, aminated graphite oxide used for CO 2 adsorption was synthesized, based on the intercalation reaction of graphite oxide (GO) with amines, including ethylenediamine (EDA), diethylenetriamine (DETA) and triethylene tetramine (TETA). The structural information, surface chemistry and thermal behavior of the adsorbent samples were characterized by X-ray powder diffraction (XRD), infrared spectroscopy (IR), transmission electron microscope (TEM), elemental analysis, particle size analysis, nitrogen adsorption as well as differential thermal and thermogravimetric analysis (DSC-TGA). CO 2 capture was investigated by dynamic adsorption experiments with N 2 -CO 2 mixed gases at 30 °C. The three kinds of graphite oxide samples modified by excess EDA, DETA and TETA showed similar adsorption behaviors seen from their breakthrough curves. Among them, the sample aminated by EDA exhibited the highest adsorption capacity with the longest breakthrough time of CO 2 . Before saturation, its adsorption capacity was up to 53.62 mg CO 2 /g sample. In addition, graphite oxide samples modified by different amount of EDA (EDA/GO raw ratio 10 wt%, 50 wt% and 100 wt%) were prepared in the ethanol. Their CO 2 adsorption performance was investigated. The experimental results demonstrated that graphite oxide with 50 wt% EDA had the largest adsorption capacity 46.55 mg CO 2 /g sample.

Nonlinear seismic analysis of a graphite reactor core

International Nuclear Information System (INIS)

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

1988-01-01

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Investigation of tungsten coatings on graphite and CFC

International Nuclear Information System (INIS)

Neu, R; Maier, H; Gauthier, E; Greuner, H; Hirai, T; Hopf, Ch; Likonen, J; Maddaluno, G; Matthews, G F; Mitteau, R; Philipps, V; Piazza, G; Ruset, C

2007-01-01

In the frame of JET's ITER-like wall (ILW) project tungsten coatings on carbon fibre reinforced carbon substrates will be used in the divertor and highly loaded areas in the main chamber. Fourteen different types of samples were produced by physical or chemical vapour deposition and vacuum plasma spray (VPS) with coating thickness of 4, 10 and 200 μm. Similarly, three different VPS W coatings (200 μm) on two different graphite substrates, were produced for use at the strike-point regions of ASDEX Upgrade. All coatings were subjected to thermal screening and thermal cycling tests in the ion beam facility GLADIS. Additionally, the coatings intended for the ILW project were exposed to edge localized mode (ELM)-like thermal loads in the electron beam facility JUDITH. A general failure mode with the CFC substrate is crack formation upon cool-down, whereas the coatings on graphite do not show any crack formation. Additionally, metallographic investigations, x-ray diffraction measurements, adhesion testing as well as measurements on the contents of light impurities were performed

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.

High-temperature annealing of graphite: A molecular dynamics study

Science.gov (United States)

Petersen, Andrew; Gillette, Victor

2018-05-01

A modified AIREBO potential was developed to simulate the effects of thermal annealing on the structure and physical properties of damaged graphite. AIREBO parameter modifications were made to reproduce Density Functional Theory interstitial results. These changes to the potential resulted in high-temperature annealing of the model, as measured by stored-energy reduction. These results show some resemblance to experimental high-temperature annealing results, and show promise that annealing effects in graphite are accessible with molecular dynamics and reactive potentials.

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.

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.

Thermal neutron dose calculation in synovium membrane for BNCS

International Nuclear Information System (INIS)

Abdalla, Khalid; Naqvi, A.A.; Maalej, N.; El-Shahat, B.

2006-01-01

A D(d,n) reaction based setup has been optimized for Boron Neutron Capture Synovectomy (BNCS). The polyethylene moderator and graphite reflector sizes were optimized to deliver the highest ratio of thermal to fast neutron yield. The neutron dose was calculated at various depths in a knee phantom loaded with boron to determine therapeutic ratios of synovium dose/skin dose and synovium dose/bone dose. Normalized to same boron loading in synovium, the values of the therapeutic ratios obtained in the present study are 12-30 times higher than the published values. (author)

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

International Nuclear Information System (INIS)

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

2011-01-01

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

"Bricks and mortar" self-assembly approach to graphitic mesoporous carbon nanocomposites

Energy Technology Data Exchange (ETDEWEB)

Fulvio, P. F.; Mayes, R.; Wang, X. Q.; Mahurin, S., M.; Bauer, J. C.; Presser, V.; McDonough, J.; Gogotsi, Y.; Dai, S.

2011-04-20

Mesoporous carbon materials do not have sufficient ordering at the atomic scale to exhibit good electronic conductivity. To date, mesoporous carbons having uniform mesopores and high surface areas have been prepared from partially-graphitizable precursors in the presence of templates. High temperature thermal treatments above 2000 °C, which are usually required to increase conductivity, result in a partial or total collapse of the mesoporous structures and reduced surface areas induced by growth of graphitic domains, limiting their applications in electric double layer capacitors and lithium-ion batteries. In this work, we successfully implemented a “brick-and-mortar” approach to obtain ordered graphitic mesoporous carbon nanocomposites with tunable mesopore sizes below 850 °C without using graphitization catalysts or high temperature thermal treatments. Phenolic resin-based mesoporous carbons act as mortar to highly conductive carbon blacks and carbon onions (bricks). The capacitance and resistivity of final materials can be tailored by changing the mortar to brick ratios.

Thermal-hydraulic modeling of porous bed reactors

International Nuclear Information System (INIS)

Araj, K.J.; Nourbakhsh, H.P.

1987-01-01

Optimum design of nuclear reactor core requires an iterative approach between the thermal-hydraulic, neutronic and operational analysis. This paper concentrates on the thermal-hydraulic behavior of a hydrogen cooled, small particle bed reactor (PBR). The PBR core, modeled here, consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flow, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit, to a common plenum. 5 refs., 1 fig., 2 tabs

Thermal comfort, physiological responses and performance of elderly during exposure to a moderate temperature drift

NARCIS (Netherlands)

Schellen, L.; Marken Lichtenbelt, van W.D.; Loomans, M.G.L.C.; Frijns, A.J.H.; Toftum, J.; Wit, de M.H.

2009-01-01

The objective of this research was to study the effects of ageing and a moderate temperature drift on human thermal comfort, physiological responses and performance. A climate room set-up with experimental subjects in the age 67-73 was used to examine the effect of a moderate temperature ramp. Eight

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.

Thermal properties of carbon inverse opal photonic crystals

International Nuclear Information System (INIS)

Aliev, Ali E.; Lee, Sergey B.; Baughman, Ray H.; Zakhidov, Anvar A.

2007-01-01

The thermal conductivity of thin-wall glassy carbon and graphitic carbon inverse opals, fabricated by templating of silica opal has been measured in the temperature range 10-400 K using transient pulse method. The heat flow through 100 A-thick layers of graphite sheets tiled on spherical surfaces of empty overlapping spheres arrayed in face-centered-cubic lattices has been analyzed in term of anisotropy factor. Taking into account high anisotropy factor in graphite, γ=342, we found that the thermal conductivity of inverse opal prepared by chemical vapor deposition infiltration is limited by heat flow across the graphitic layers in bottleneck, κ-perpendicular =3.95 W/m K. The electronic contribution to the thermal conductivity, κ e(300K) =3.7x10 -3 W/m K is negligible comparing to the measured value, κ (300K) =0.33 W/m K. The obtained phonon mean free path, l=90 nm is comparable with the graphite segments between hexagonal array of interconnections

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.

Studies on synthesis and hydrogenation behaviour of graphitic nanofibres prepared through palladium catalyst assisted thermal cracking of acetylene

International Nuclear Information System (INIS)

Gupta, Bipin Kumar; Tiwari, R.S.; Srivastava, O.N.

2004-01-01

The nano-variants of carbon including graphitic nanofibres (GNF) have recently been considered to be exotic (light weight, high storage capacity) hydrogen storage materials. In the present paper, we report growth of aligned bundles of GNF. The length and width of cross-section of the bundles is ∼50 and ∼25 μm, respectively. The length of individual GNF is ∼50 μm and diameter ∼0.25 μm. The GNFs have been synthesized through thermal decomposition of acetylene using palladium (Pd) sheets as catalyst. This represents a new form of catalyst. The GNFs bundles grown by the present method are easier to hydrogenate. They adsorb hydrogen at a lower pressure of ∼80 atm as against ∼120 atm for the GNF grown in the earlier studies. The storage capacity obtained in the present investigation is ∼17 wt.%. Electron microscopic investigations reveal that as against the as grown GNF, the hydrogenated version embodies microstructures exhibiting fragmentation of graphitic layer bundles. The reasons for the growth of GNF in the form of aligned bundles, the ease of hydrogenation and relevance of GNF fragmentation after hydrogenation have been outlined

Studies on synthesis and hydrogenation behaviour of graphitic nanofibres prepared through palladium catalyst assisted thermal cracking of acetylene

Energy Technology Data Exchange (ETDEWEB)

Gupta, Bipin Kumar; Tiwari, R.S.; Srivastava, O.N

2004-11-03

The nano-variants of carbon including graphitic nanofibres (GNF) have recently been considered to be exotic (light weight, high storage capacity) hydrogen storage materials. In the present paper, we report growth of aligned bundles of GNF. The length and width of cross-section of the bundles is {approx}50 and {approx}25 {mu}m, respectively. The length of individual GNF is {approx}50 {mu}m and diameter {approx}0.25 {mu}m. The GNFs have been synthesized through thermal decomposition of acetylene using palladium (Pd) sheets as catalyst. This represents a new form of catalyst. The GNFs bundles grown by the present method are easier to hydrogenate. They adsorb hydrogen at a lower pressure of {approx}80 atm as against {approx}120 atm for the GNF grown in the earlier studies. The storage capacity obtained in the present investigation is {approx}17 wt.%. Electron microscopic investigations reveal that as against the as grown GNF, the hydrogenated version embodies microstructures exhibiting fragmentation of graphitic layer bundles. The reasons for the growth of GNF in the form of aligned bundles, the ease of hydrogenation and relevance of GNF fragmentation after hydrogenation have been outlined.

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

Transition between scanning tunneling microscopy images of alkane derivatives on graphite

International Nuclear Information System (INIS)

Hibino, Masahiro; Tsuchiya, Hiroshi

2015-01-01

Graphical abstract: - Highlights: • SAMs of dialkyl sulfides form at the liquid–graphite interface. • STM contrast of molecules change reversibly between zigzag and aligned bright spot patterns. • The free energy for contrast change is smaller than the thermal energy (RT). • STM contrast change is caused by electronic effects and registry of the alkyl chains. - Abstract: Self-assembled monolayers of alkylated sulfides containing two alkyl chains and a sulfur atom positioned at the center of the molecules were studied on a graphite surface using scanning tunneling microscopy (STM). STM images of the closed-packed alkyl chains that extend linearly from the sulfur atoms change reversibly between a zigzag pattern and an aligned bright spot pattern on a time scale of minutes. The observation times of the zigzag and aligned bright spot patterns indicate that the difference between the free energies of these two stable molecular configurations with respect to the graphite surface is smaller than their thermal energies in the presence of a solvent, and 10 times smaller than the theoretical free energy between parallel and perpendicular configurations of the alkyl chains on graphite under vacuum. The change in the contrast of the STM images occurred owing to the electronic effects that depend on the registry of the alkyl chains on the graphite surface, and not by the classical observation of transfer between parallel and perpendicular orientations of alkyl chains on the surface.

Tailoring thermal transport properties of graphene by nitrogen doping

Energy Technology Data Exchange (ETDEWEB)

Zhang, Tingting; Li, Jianhua; Cao, Yuwei; Zhu, Liyan, E-mail: lyzhu@hytc.edu.cn; Chen, Guibin, E-mail: gbchen@hytc.edu.cn [Huaiyin Normal University, School of Physics and Electronic & Electrical Engineering (China)

2017-02-15

The influence of two different nitrogen doping configurations, graphite-like and pyridinic-like nitrogen doping (denoted as graphite-N and pyridinic-N hereafter, respectively), on the thermal conduction of graphene is carefully studied via non-equilibrium molecular dynamic (NEMD) simulations. The thermal conductivity is more strongly suppressed in the pyridinic-N-doped graphene than that in the graphite-N-doped sample, which can be well understood from the changes in bond strength between nitrogen and carbon atoms, phonon group velocities, phonon density of states, participation ratio, and phonon transmission. Our study indicates that the pyridinic-N doping is an efficient method to tune the thermal conduction in graphene, especially for the situation where low thermal conductivity is requested, e.g., thermoelectric applications and thermal shielding.

Subsurface methane formation in graphite due to exposure to H+ and D+

International Nuclear Information System (INIS)

Chiu, S.; Haasz, A.A.

1994-01-01

The extent of H-D mixing in the form of mixed-isotope methane formation during simultaneous H + /D + bombardment of graphite was measured and found to be a sensitive function of H + -D + ion range separation. The result strongly supports the model that methane molecules are formed at the end of ion range in the bulk of graphite. A long steadily decreasing transient was observed in the methane emission when bombarding a virgin graphite sample with 10 keV D + at 800 K. The effect was less pronounced for low density fine grain isographites (EK98) than for high density pyrolytic graphite (HPG99) and monocrystal carbons. We propose a model which attributes this transient to the creation of new internal ''surfaces'', formed during ion irradiation along with ion-created micropaths. The increase in internal surface area effectively ''dilutes'' the inner surface hydrogen concentration in the region where incident hydrogen ions thermalize. We propose that, initially, methane molecules are formed mainly on inherent internal surfaces (pore and grain/crystallite boundary surfaces) in the region where hydrogen ions thermalize; then as damage builds up, newly produced internal ''surfaces'' also contribute to methane formation. (orig.)

New thermal neutron calibration channel at LNMRI/IRD

International Nuclear Information System (INIS)

Astuto, A.; Lopes, R.T.; Patrao, K.C.S.; Fonseca, E.S.; Pereira, W.W.

2015-01-01

A new standard thermal neutron flux unit was designed in the National Ionizing Radiation Metrology Laboratory (LNMRI) for calibration of neutron detectors. Fluence is achieved by moderation of four 241 Am-Be sources with 596 GBq each, in a facility built with graphite and paraffin blocks. The study was divided into two stages. First, simulations were performed using MCNPX code in different geometric arrangements, seeking the best performance in terms of fluence and their uncertainties. Last, the system was assembled based on the results obtained on the simulations. The simulation results indicate quasi-homogeneous fluence (less than 1%) in the central chamber. (author)

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.

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)

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

Science.gov (United States)

Sadeghifar, Hamidreza; Djilali, Ned; Bahrami, Majid

2015-01-01

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

SEM/XPS analysis of fractured adhesively bonded graphite fibre surface resin-rich/graphite fibre composites

Science.gov (United States)

Devilbiss, T. A.; Wightman, J. P.; Progar, D. J.

1988-01-01

Samples of graphite fiber-reinforced polyimide were fabricated allowing the resin to accumulate at the composite surface. These surface resin-rich composites were then bonded together and tested for lap shear strength both before and after thermal aging. Lap shear strength did not appear to show a significant improvement over that previously recorded for resin-poor samples and was shown to decrease with increasing aging time and temperature.

The irradiation creep characteristics of graphite to high fluences

International Nuclear Information System (INIS)

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

1988-01-01

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

Feasibility study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

International Nuclear Information System (INIS)

Ohnuki, A.; Kureta, M.; Liu, W.; Tamai, H.; Akimoto, H.

2004-01-01

Research and development project for investigating thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured light-water reactor technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important issues for the RMWR because of the tight-lattice configuration. The project has mainly consisted of a large-scale thermal-hydraulic test and development of analytical methods named modeling engineering. In the large-scale test, 37-rod bundle experiments can be performed. Steady-state critical power experiments have been achieved in the test facility and the experimental data reveal the feasibility of RMWR

Temperature-dependent thermal properties of a paraffin phase change material embedded with herringbone style graphite nanofibers

International Nuclear Information System (INIS)

Warzoha, Ronald J.; Weigand, Rebecca M.; Fleischer, Amy S.

2015-01-01

Highlights: • The thermal properties of a PCM with nanofibers are determined. • The solid-phase thermal conductivity scales exponentially with volume fraction. • The liquid-phase thermal conductivity is only enhanced beyond a critical percolation threshold. • The nanoscale interface resistance depends on the nanoparticle’s dimensionality. • The thermal diffusivity and volumetric heat capacity of the nanoenhanced PCMs are found. - Abstract: In many studies, carbon nanoparticles with high values of thermal conductivity (10–3000 W/m K) have been embedded into phase change thermal energy storage materials (PCMs) in order to enhance their bulk thermal properties. While a great deal of work to date has focused on determining the effect of these nanoparticles on a PCM’s solid phase thermal properties, little is known about their effect on its liquid phase thermal properties. Thus, in this study, the effect of implanting randomly oriented herringbone style graphite nanofibers (HGNF, average diameter = 100 nm, average length = 20 μm) on the bulk thermal properties of an organic paraffin PCM (IGI 1230A, T melt = 329.15 K) in both the solid and liquid phase is quantified. The bulk thermal conductivity, volumetric heat capacity and thermal diffusivity of HGNF/PCM nanocomposites are obtained as a function of temperature and HGNF volume loading level. It is found that the property enhancement varies significantly depending on the material phase. In order to explain the difference between solid and liquid phase thermal properties, heat flow at the nanoparticle–PCM and nanoparticle–nanoparticle interfaces is examined as a function of HGNF loading level and temperature. To do this, the solid and liquid phase thermal boundary resistances (TBRs) between the nanoparticles and the surrounding PCM and/or between contacting nanoparticles are found. Results suggest that the TBR at the HGNF–PCM interface is nearly double the TBR across the HGNF–HGNF interface in

Braze Development of Graphite Fiber for Use in Phase Change Material Heat Sinks

Science.gov (United States)

Quinn, Gregory; Beringer, Woody; Gleason, Brian; Stephan, Ryan

2011-01-01

Hamilton Sundstrand (HS), together with NASA Johnson Space Center, developed methods to metallurgically join graphite fiber to aluminum. The goal of the effort was to demonstrate improved thermal conductance, tensile strength and manufacturability compared to existing epoxy bonded techniques. These improvements have the potential to increase the performance and robustness of phase change material heat sinks that use graphite fibers as an interstitial material. Initial work focused on evaluating joining techniques from four suppliers, each consisting of a metallization step followed by brazing or soldering of one inch square blocks of Fibercore graphite fiber material to aluminum end sheets. Results matched the strength and thermal conductance of the epoxy bonded control samples, so two suppliers were down-selected for a second round of braze development. The second round of braze samples had up to a 300% increase in strength and up to a 132% increase in thermal conductance over the bonded samples. However, scalability and repeatability proved to be significant hurdles with the metallization approach. An alternative approach was pursued which used a nickel braze allow to prepare the carbon fibers for joining with aluminum. Initial results on sample blocks indicate that this approach should be repeatable and scalable with good strength and thermal conductance when compared with epoxy bonding.

Temperature control of the graphite stack of the reactor RBMK-1500

International Nuclear Information System (INIS)

Lesnoj, S.

1998-01-01

The paper includes general information about RBMK-1500 reactor, construction features and main technical data; graphite moderator stack, temperature channel, thermocouple TXA-1379, its basic technical and metrologic parameters as well as its advantages and disadvantages