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Sample records for nuclear graphite properties

  1. IAEA international database on irradiated nuclear graphite properties

    International Nuclear Information System (INIS)

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

    2000-02-01

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

  2. Characterization of nuclear graphite elastic properties using laser ultrasonic methods

    Science.gov (United States)

    Zeng, Fan W.; Han, Karen; Olasov, Lauren R.; Gallego, Nidia C.; Contescu, Cristian I.; Spicer, James B.

    2015-05-01

    Laser ultrasonic methods have been used to characterize the elastic behaviors of commercially-available and legacy nuclear graphites. Since ultrasonic techniques are sensitive to various aspects of graphite microstructure including preferred grain orientation, microcrack orientation and porosity, laser ultrasonics is a candidate technique for monitoring graphite degradation and structural integrity in environments expected in high-temperature, gas-cooled nuclear reactors. Aspects of materials texture can be assessed by studying ultrasonic wavespeeds as a function of propagation direction and polarization. Shear wave birefringence measurements, in particular, can be used to evaluate elastic anisotropy. In this work, laser ultrasonic measurements of graphite moduli have been made to provide insight into the relationship between the microstructures and the macroscopic stiffnesses of these materials. In particular, laser ultrasonic measurements have been made using laser line sources to produce shear waves with specific polarizations. By varying the line orientation relative to the sample, shear wave birefringence measurements have been recorded. Results from shear wave birefringence measurements show that an isostatically molded graphite, such as PCIB, behaves isotropically, while an extruded graphite, such as H-451, displays significant ultrasonic texture. Graphites have complicated microstructures that depend on the manufacturing processes used, and ultrasonic texture in these materials could originate from grain orientation and preferred microcrack alignment. Effects on material isotropy due to service related microstructural changes are possible and the ultimate aim of this work is to determine the degree to which these changes can be assessed nondestructively using laser ultrasonics measurements.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-15

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

  4. 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

  5. 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)

  6. Leveraging comprehensive baseline datasets to quantify property variability in nuclear-grade graphites

    Energy Technology Data Exchange (ETDEWEB)

    Carroll, Mark C., E-mail: mark.carroll@inl.gov [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-2213 (United States); Windes, William E.; Rohrbaugh, David T. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-2213 (United States); Strizak, Joseph P.; Burchell, Timothy D. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6088 (United States)

    2016-10-15

    Highlights: • An effort is underway to fully quantify the properties of nuclear-grade graphites. • Physical and mechanical properties of graphite are best characterized by distributions. • The Weibull distribution is most representative of graphite based on goodness-of-fit. • Fine-grained isomolded grades exhibit higher Weibull modulus values, indicative of more homogeneous properties. - Abstract: The full characterization of the physical and mechanical properties of candidate nuclear-grade graphites is highly dependent upon an understanding of the distribution of values that are inherent to graphite. Not only do the material properties of graphites vary considerably between grades owing to the raw materials sources, filler particle type and size, methods of compaction, and production process parameters, but variability is observed between billets of the same grade from a single batch and even across spatial positions within a single billet. Properly enveloping the expected properties of interest requires both a substantial amount of data to statistically capture this variability and a representative distribution capable of accurately describing the range of values. A two-parameter Weibull distribution is confirmed to be representative of the distribution of physical (density, modulus) and mechanical (compressive, flexure, and tensile strength) values in five different nuclear-grades of graphite. The fine-grained isomolded grades tend toward higher Weibull modulus and characteristic strength values, while the extruded grade being examined exhibits relatively large distributions in property values. With the number of candidate graphite specimens that can undergo full irradiation exposure and subsequent testing having limited feasibility with regard to economics and timely evaluations, a proper capture of the raw material variability in an unirradiated state can provide crucial supplementary resolution to the limited amount of available data on irradiated

  7. 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)

  8. Graphite in Science and Nuclear Technique

    OpenAIRE

    Zhmurikov, E. I.; Bubnenkov, I. A.; Dremov, V. V.; Samarin, S. I.; Pokrovsky, A. S.; Harkov, D. V.

    2013-01-01

    The monograph is devoted to the application of graphite and graphite composites in science and technology. The structure and electrical properties, the technological aspects of production of high-strength synthetic graphites, the dynamics of the graphite destruction, traditionally used in the nuclear industry are discussed. It is focuses on the characteristics of graphitization and properties of graphite composites based on carbon isotope 13C. The book is based, generally, on the original res...

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

    International Nuclear Information System (INIS)

    Wickham, A.J.; Humbert, D.

    2006-06-01

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

  10. Graphite in Science and Nuclear Technology

    OpenAIRE

    Zhmurikov, Evgenij

    2015-01-01

    This review is devoted to the application of graphite and graphite composites in the science and technology. Structure and electrical properties, technological aspects of producing of high-strength artificial graphite and dynamics of its destruction are considered. These type of graphite are traditionally used in the nuclear industry, so author concentrates on actual problems of application and testing of graphite materials in modern science and technology. Translated from chapters 1 of monog...

  11. 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)

  12. IAEA International Database on Irradiated Nuclear Graphite Properties. 7th meeting of the Technical Steering Committee

    International Nuclear Information System (INIS)

    Humbert, D.; Wickham, A.J.

    2005-06-01

    This report summarizes the Consultant Meeting '7th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties' held on 16-17 March 2005 at the IAEA Headquarters, Vienna, Austria. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database and to make recommendations for actions for the next year. The purposes of the meeting were fully met. This report contains the current status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

  13. Summary report of consultants' meeting on IAEA international database on irradiated nuclear graphite properties

    International Nuclear Information System (INIS)

    Humbert, D.; Wickham, A.J.

    2007-06-01

    The '9th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties' was held on 26-27 March 2007 at the IAEA Headquarters, Vienna, Austria. All discussions, recommendations and actions of this Consultants' Meeting are recorded in this report. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database and make recommendations for actions for the next year. This report contains the current status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

  14. IAEA International Database on Irradiated Nuclear Graphite Properties. 6th meeting of the Technical Steering Committee

    International Nuclear Information System (INIS)

    Humbert, D.; Wickham, A.J.

    2004-12-01

    This report summarizes the Consultant Meeting 6th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties' held on 16-17 September 2004 at Plas Tan-Y-Bwlch, Maentwrog, Gwynedd, UK. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database and to make recommendations for actions for the next year. The purposes of the meeting were fully met. This report contains the current status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  16. Production of nuclear graphite in France

    International Nuclear Information System (INIS)

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

    1955-01-01

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

  17. The influence of roughness on tribological properties of nuclear grade graphite

    International Nuclear Information System (INIS)

    Luo Xiaowei; Yu Suyuan; Sheng Xuanyu; He Shuyan

    2006-01-01

    The influence of surface roughness on tribological properties of graphite IG-11 was investigated on a standard SRV tester. The experimental condition was selected as: 30 N normal load, room temperature and a 10 Hz frequency with different strokes. The experiments environments included helium and air. Five types of roughness were studied in the experiments. The experiments revealed that the surface roughness greatly affected the graphite friction behavior. When the friction surface was smooth, the friction coefficient was high because of intensive adhesion accompanied by many pits at the friction surface. When the friction surface was rough, the adhesion was very poor, but the wear was excessive and generated many graphite particles at the friction surface. These particles can separate the friction surfaces, which reduced the friction action between them. For very rough specimens, the friction coefficient decreased with sliding velocity at about 0.004 m/s and then increases gradually

  18. 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

  19. IAEA International Database on Irradiated Nuclear Graphite Properties. Summary report of consultants' meeting. 12. meeting of the Technical Steering Committee

    International Nuclear Information System (INIS)

    Chung, H.K.; Wickham, A.J.

    2010-02-01

    The 12th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties was held on 12-13 November 2009 at the IAEA Headquarters, Vienna, Austria. All discussions, recommendations and actions of this Consultants' Meeting are recorded in this report. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database, and make recommendations for action over the next year. This report contains the status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

  20. 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

  1. Oxidation behavior of IG and NBG nuclear graphites

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-01-15

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

  2. Nondestructive evaluation of nuclear-grade graphite

    Science.gov (United States)

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

    2012-05-01

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

  3. Electronic properties of graphite

    International Nuclear Information System (INIS)

    Schneider, J.

    2010-10-01

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

  4. Cementation of Nuclear Graphite Using Geopolymers

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  5. 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)

  6. Improving molten fluoride salt and Xe135 barrier property of nuclear graphite by phenolic resin impregnation process

    Science.gov (United States)

    He, Zhao; Lian, Pengfei; Song, Yan; Liu, Zhanjun; Song, Jinliang; Zhang, Junpeng; Feng, Jing; Yan, Xi; Guo, Quangui

    2018-02-01

    A densification process has been conducted on isostatic graphite (IG-110, TOYO TANSO CO., Ltd., Japan) by impregnating phenolic resin to get the densified isostatic graphite (D-IG-110) with pore diameter of nearly 11 nm specifically for molten salt reactor application. The microstructure, mechanical, thermophysical and other properties of graphite were systematically investigated and compared before and after the densification process. The molten fluoride salt and Xe135 penetration in the graphite were evaluated in a high-pressure reactor and a vacuum device, respectively. Results indicated that D-IG-110 exhibited improved properties including infiltration resistance to molten fluoride salt and Xe135 as compared to IG-110 due to its low porosity of 2.8%, the average pore diameter of 11 nm and even smaller open pores on the surface of the graphite. The fluoride salt infiltration amount of IG-110 was 13.5 wt% under 1.5 atm and tended to be saturated under 3 atm with the fluoride salt occupation of 14.8 wt%. As to the D-IG-110, no salts could be detected even up to 10 atm attempted loading. The helium diffusion coefficient of D-IG-110 was 6.92 × 10-8 cm2/s, significantly less than 1.21 × 10-2 cm2/s of IG-110. If these as-produced properties for impregnated D-IG-110 could be retained during MSR operation, the material could prove effective at inhibiting molten fluoride salt and Xe135 inventories in the graphite.

  7. Modelling property changes in graphite irradiated at changing irradiation temperature

    CSIR Research Space (South Africa)

    Kok, S

    2011-01-01

    Full Text Available A new method is proposed to predict the irradiation induced property changes in nuclear; graphite, including the effect of a change in irradiation temperature. The currently used method; to account for changes in irradiation temperature, the scaled...

  8. The Fracture Toughness of Nuclear Graphites Grades

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-01

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

  9. On estimating the fracture probability of nuclear graphite components

    International Nuclear Information System (INIS)

    Srinivasan, Makuteswara

    2008-01-01

    The properties of nuclear grade graphites exhibit anisotropy and could vary considerably within a manufactured block. Graphite strength is affected by the direction of alignment of the constituent coke particles, which is dictated by the forming method, coke particle size, and the size, shape, and orientation distribution of pores in the structure. In this paper, a Weibull failure probability analysis for components is presented using the American Society of Testing Materials strength specification for nuclear grade graphites for core components in advanced high-temperature gas-cooled reactors. The risk of rupture (probability of fracture) and survival probability (reliability) of large graphite blocks are calculated for varying and discrete values of service tensile stresses. The limitations in these calculations are discussed from considerations of actual reactor environmental conditions that could potentially degrade the specification properties because of damage due to complex interactions between irradiation, temperature, stress, and variability in reactor operation

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1955-07-01

    The graphite intended for the construction of the reactors is obtained by the usual process: confection of a cake from coke of oil and tar, cooked (in a electric oven) then the product of cook is graphitized, also by electric heating. The use of the air transportation and the control of conditions cooking and graphitization have permitted to increase the nuclear graphite production as well as to better control their physical and mechanical properties and to reduce to the minimum the unwanted stains. (M.B.) [French] Le graphite destine a la construction des reacteurs est obtenu par le procede usuel: confection d'une pate a partir de coke de petrole et de brai, cuisson de cette pate (au four electrique) puis graphitation du produit cuit, egalement par chauffage electrique. L'usage du transport pneumatique et le controle des conditions cuisson et de graphitation ont permit d'augmenter la production de graphite nucleaire ainsi que de mieux controler ses proprietes physiques et mecaniques et de reduire au minimum les souillures accidentelles. (M.B.)

  11. 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)

  12. 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

  13. 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)

  14. Summary report of consultants meeting on IAEA International Database on Irradiated Nuclear Graphite Properties. 11. meeting of the Technical Steering Committee

    International Nuclear Information System (INIS)

    Humbert, D.; Wickham, A.J.

    2009-05-01

    The 11th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties was held on 25-26 March 2009 at the IAEA Headquarters, Vienna, Austria. All discussions, recommendations and actions of this Consultants' Meeting are recorded in this report. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database, and make recommendations for action over the next year. This report contains the status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

  15. Summary report of consultants' meeting - IAEA International Database on Irradiated Nuclear Graphite Properties. 8th meeting of the Technical Steering Committee

    International Nuclear Information System (INIS)

    Humbert, D.; Wickham, A.J.

    2006-05-01

    The '8th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties' was held on 15-16 March 2006 at the IAEA Headquarters, Vienna, Austria. All discussions, recommendations and actions of this Consultants' Meeting are recorded in this report. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database and make recommendations for actions for the next year. This report contains the current status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

  16. Nuclear graphite for high temperature reactors

    International Nuclear Information System (INIS)

    Marsden, B.J.

    2001-01-01

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

  17. Vapour pressure of caesium over nuclear graphite

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  18. 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

  19. 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

  20. Elastic properties of graphite and interstitial defects

    International Nuclear Information System (INIS)

    Ayasse, J.-B.

    1977-01-01

    The graphite elastic constants C 33 and C 44 , reflecting the interaction of the graphitic planes, were experimentally measured as a function of irradiation and temperature. A model of non-central strength atomic interaction was established to explain the experimental results obtained. This model is valid at zero temperature. The temperature dependence of the elastic properties was analyzed. The influence of the elastic property variations on the specific heat of the lattice at very low temperature was investigated [fr

  1. Direct reading spectrochemical analysis of nuclear graphite

    International Nuclear Information System (INIS)

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

    1964-01-01

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

  2. 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

  3. 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

  4. 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

  5. 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)

  6. Thermal Properties of G-348 Graphite

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-01

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

  7. 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.

  8. Spectroscopical determination of impurities in nuclear graphite

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  9. Comparison of Oxidation Characteristics of Selected Nuclear Graphite Grades

    International Nuclear Information System (INIS)

    Chi, Se Hwan; Kim, Gen Chan

    2010-02-01

    The oxidation behavior of some selected nuclear graphite grades (i.e., IG-110, IG-430, NBG-18, NBG-25) were compared in view of their filler coke type and the physical property of the grades. Oxidation rates were determined at six temperatures between 600 ∼ 960 .deg. C in air by using a three-zone vertical tube furnace at a 10 L/min air flow rate. The specimens were a cylinder with a 25.4 mm diameter and a 25.4 mm length. Results showed that, even though the four examined nuclear graphite grades showed a highly temperature-sensitive oxidation behavior through out the test temperature range of 600 ∼ 950 .deg. C, the differences between the grades were not significant. The oxidation rates determined for a 5∼10 % weight loss at the six temperatures were nearly the same except for 702 and 808 .deg. C, where the pitch coke graphites showed an apparent decrease in their oxidation rate, more so than the petroleum coke graphites. These effects of the coke type reduced or nearly disappeared with an increasing temperature. The average activation energy determined for 608 ∼ 808 .deg. C was 161.5 ± 7.3 kJ/mol, showing that the dominant oxidation reaction occurred by a chemical control

  10. 77 FR 51581 - Request for a License To Export Nuclear Grade Graphite

    Science.gov (United States)

    2012-08-24

    ... NUCLEAR REGULATORY COMMISSION Request for a License To Export Nuclear Grade Graphite Pursuant to... 27, 2012, graphite for of nuclear grade graphite to the XMAT424, 11006032. nuclear end use. graphite. Shanghai Institute of Applied Physics in China to test various types of nuclear grade graphite material in...

  11. Electrical properties of Egyptian natural graphite

    International Nuclear Information System (INIS)

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

    1992-06-01

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

  12. 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

  13. 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.

  14. Testing of Small Graphite Samples for Nuclear Qualification

    Energy Technology Data Exchange (ETDEWEB)

    Julie Chapman

    2010-11-01

    Accurately determining the mechanical properties of small irradiated samples is crucial to predicting the behavior of the overal irradiated graphite components within a Very High Temperature Reactor. The sample size allowed in a material test reactor, however, is limited, and this poses some difficulties with respect to mechanical testing. In the case of graphite with a larger grain size, a small sample may exhibit characteristics not representative of the bulk material, leading to inaccuracies in the data. A study to determine a potential size effect on the tensile strength was pursued under the Next Generation Nuclear Plant program. It focuses first on optimizing the tensile testing procedure identified in the American Society for Testing and Materials (ASTM) Standard C 781-08. Once the testing procedure was verified, a size effect was assessed by gradually reducing the diameter of the specimens. By monitoring the material response, a size effect was successfully identified.

  15. 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

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

    Science.gov (United States)

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

    2018-05-01

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

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

    International Nuclear Information System (INIS)

    Johnson, P.A.V.

    1982-01-01

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

  18. A systematic study of acoustic emission from nuclear graphites

    International Nuclear Information System (INIS)

    Neighbour, G.B.; McEnaney, B.

    1996-01-01

    Acoustic emission (AE) monitoring has been identified as a possible method to determine internal stresses in nuclear graphites using the Kaiser effect, i.e., on stressing a graphite that has been subject to a prior stress, the onset of AE occurs at the previous peak stress. For three nuclear graphites (PGA, IM1-24 and VNEC), AE was monitored during both monotonic and cyclic loading to failure in tensile, compressive and flexural test modes. For unirradiated graphites, the Kaiser effect was not found in cyclic loading, but a Felicity effect was observed, i.e., the onset of AE occurred below the previously applied peak stress. The Felicity effect was attributed to time-dependent relaxation and recovery processes and was characterized using a new parameter, the Recovery ratio. It was shown that AE can be used to monitor creep strain and creep recovery in graphites at zero load. The AE-time responses from these experiments were fitted to equations similar to those used for creep strain-time at elevated temperatures. The number of AE counts from irradiated graphites were greater than those from unirradiated graphites, subject to similar stresses, due to increases in porosity caused by radiolytic oxidation. A Felicity effect was also observed on cyclic loading of irradiated graphites, but no evidence for a Kaiser effect was found for irradiated graphites loaded monotonically to failure. Thus internal stresses in irradiated graphites could not be measured using AE. This was attributed to relaxation and recovery processes that occur between removing the irradiated graphite from the reactor and AE testing. This work indicated that AE monitoring is not a suitable technique for measuring internal stresses in irradiated graphite. (author). 19 refs, 6 figs, 6 tabs

  19. Cementation of nuclear graphite using geo-polymers

    International Nuclear Information System (INIS)

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

    2012-01-01

    Geo-polymers are solid aluminosilicate materials usually formed by alkali hydroxide or alkali silicate activation of solid precursors such as coal fly ash, calcined clay and/or metallurgical slag. Today the primary application of geo-polymer technology is in the development of alternatives to Portland-based cements. Variations in the ratio of aluminium to silicon, and alkali to silicon or addition of structure support, produce geo-polymers with different physical and mechanical properties. These materials have an amorphous three-dimensional structure that gives geo-polymers certain properties, such as fire and acid resistance, low leach rate, which make them an ideal substitute for ordinary Portland cement (OPC) in a wide range of applications especially in conditioning and storage of radioactive waste. Therefore investigations have been initiated about how and to which amount graphite as a hydrophobic material can be mixed with cement or concrete to form stable waste products and which concretes fulfill the specifications at best. As result geo-polymers have been identified as a promising matrix for graphite containing nuclear wastes. With geo-polymers both favorable properties in the cementation process and a high long time structural stability of the products can be achieved. (authors)

  20. 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

  1. 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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  3. Microstructural modelling of nuclear graphite using multi-phase models

    International Nuclear Information System (INIS)

    Berre, C.; Fok, S.L.; Marsden, B.J.; Mummery, P.M.; Marrow, T.J.; Neighbour, G.B.

    2008-01-01

    This paper presents a new modelling technique using three-dimensional multi-phase finite element models in which meshes representing the microstructure of thermally oxidised nuclear graphite were generated from X-ray micro-tomography images. The density of the material was related to the image greyscale using Beer-Lambert's law, and multiple phases could thus be defined. The local elastic and non-linear properties of each phase were defined as a function of density and changes in Young's modulus, tensile and compressive strength with thermal oxidation were calculated. Numerical predictions compared well with experimental data and with other numerical results obtained using two-phase models. These models were found to be more representative of the actual microstructure of the scanned material than two-phase models and, possibly because of pore closure occurring during compression, compressive tests were also predicted to be less sensitive to the microstructure geometry than tensile tests

  4. Calculated bond properties of K adsorbed on graphite

    International Nuclear Information System (INIS)

    Hjortstam, O.; Wills, J.M.; Johansson, B.; Eriksson, O.

    1998-01-01

    The properties of the chemical bond of K adsorbed on a graphite(0001) surface have been studied for different coverages, by means of a full-potential slab method. Specific modifications of the Hamiltonian are performed in order to make it possible to study K on graphite in the disperse phase (dilute limit). It is found that K forms a metallic state when covering a graphite surface with a (2x2) coverage. For a (3x3) coverage as well as in the disperse phase K is found to form an ionic bond with graphite. It is shown that in the disperse phase, the hybridization between the K 4s level and graphite is weak. Our findings are consistent with recent experiments. Furthermore the cohesive energies of K adsorption on graphite are found to be larger in the (2x2) coverage compared to the (3x3) coverage. copyright 1998 The American Physical Society

  5. Variation of the properties of siliconized graphite during neutron irradiation

    International Nuclear Information System (INIS)

    Virgil'ev, Y.S.; Chugunova, T.K.; Pikulik, R.G.

    1986-01-01

    The authors evaluate the radiation-induced property changes in siliconized graphite of the industrial grades SG-P and SG-M. The authors simultaneously tested the reference (control) specimens of graphite that are used as the base for obtaining the SG-M siliconized graphite by impregnating with silicon. The suggested scheme (model) atributes the dimensional changes of the siliconized graphite specimens to the effect of the quantitative ratio of the carbide phase and carbon under different conditions of irradiation. If silicon is insufficient for the formation of a dense skeleton, graphite plays a devisive role, and it may be assumed that at an irradiation temperature greater than 600 K, the material shrinks. The presence of isolated carbide inclusions also affects the physicomechanical properties (including the anitfriction properties)

  6. Rules for design of nuclear graphite core components - some considerations and approaches

    International Nuclear Information System (INIS)

    Svalbonas, V.; Stilwell, T.C.; Zudans, Z.

    1978-01-01

    The use of graphite as a structural element presents unusual problems both for the designer and stress analysist. When the structure happens to be a nuclear reactor core, these problems are significantly magnified both by the environment and the attendant safety requirements. In the high temperature gas reactor (HTGR) core a large number of elements are constructed of nuclear graphite. This paper discusses the attendant difficulties, and presents some approaches, for ASME code safety-consistent design and analysis. The statistical scatter of material properties, which complicates even the definitions of allowable stress, as well as the brittle, anisotropic, inhomogeneous nature of the graphite was considered. The study of this subject was undertaken under contract to the U.S. Nuclear Regulatory Commission. (Auth.)

  7. Differences in the irradiation effects of IG-110 and IG-430 nuclear graphites : effects of coke difference

    International Nuclear Information System (INIS)

    Chi, Se Hwan; Kim, Gen Chan; Kim, Eung Seon; Hong, Jin Ki; Chang, Jong Hwa

    2005-01-01

    In the high temperature gas cooled reactors (HTGRs), graphite acts as a moderator and reflector as well as a major structural component that may provide channels for the fuel and coolant gas, channels for control and shut down, and thermal and neutron shielding. During a reactor operation, many of the physical, chemical and mechanical properties of these graphite components are significantly modified as a function of the temperature, environment, and an irradiation. On the other hand, currently, all the nuclear graphites are being manufactured from two types of cokes, i.e., petroleum and coal-tar pitch coke, and it has been understood that the type of coke plays the most critical role determining the properties of a specific graphite grade. To investigate the effects of coke types on the irradiation response of a graphite, two graphites of different cokes were irradiated by 3 MeV C+ ions and the differences in the response of ion-irradiation were investigated

  8. Comparison of the oxidation rate and degree of graphitization of selected IG and NBG nuclear graphite grades

    Science.gov (United States)

    Chi, Se-Hwan; Kim, Gen-Chan

    2008-10-01

    The oxidation rate and degree of graphitization (DOG) were determined for some selected nuclear graphite grades (i.e., IG-110, IG-430, NBG-18, NBG-25) and compared in view of their filler coke type (i.e., pitch or petroleum coke) and the physical property of the grades. Oxidation rates were determined at six temperatures between 600 and 960 °C in air by using a three-zone vertical tube furnace at a 10 l/min air flow rate. The specimens were a cylinder with a 25.4 mm diameter and a 25.4 mm length. The DOG was determined based on the lattice parameter c determined from an X-ray diffraction (XRD). Results showed that, even though the four examined nuclear graphite grades showed a highly temperature-sensitive oxidation behavior through out the test temperature range of 600-950 °C, the differences between the grades were not significant. The oxidation rates determined for a 5-10% weight loss at the six temperatures were nearly the same except for 702 and 808 °C, where the pitch coke graphites showed an apparent decrease in their oxidation rate, more so than the petroleum coke graphites. These effects of the coke type reduced or nearly disappeared with an increasing temperature. The average activation energy determined for 608-808 °C was 161.5 ± 7.3 kJ/mol, showing that the dominant oxidation reaction occurred by a chemical control. A relationship between the oxidation rate and DOG was not observed.

  9. Comparison of the oxidation rate and degree of graphitization of selected IG and NBG nuclear graphite grades

    International Nuclear Information System (INIS)

    Chi, Se-Hwan; Kim, Gen-Chan

    2008-01-01

    The oxidation rate and degree of graphitization (DOG) were determined for some selected nuclear graphite grades (i.e., IG-110, IG-430, NBG-18, NBG-25) and compared in view of their filler coke type (i.e., pitch or petroleum coke) and the physical property of the grades. Oxidation rates were determined at six temperatures between 600 and 960 deg. C in air by using a three-zone vertical tube furnace at a 10 l/min air flow rate. The specimens were a cylinder with a 25.4 mm diameter and a 25.4 mm length. The DOG was determined based on the lattice parameter c determined from an X-ray diffraction (XRD). Results showed that, even though the four examined nuclear graphite grades showed a highly temperature-sensitive oxidation behavior through out the test temperature range of 600-950 deg. C, the differences between the grades were not significant. The oxidation rates determined for a 5-10% weight loss at the six temperatures were nearly the same except for 702 and 808 deg. C, where the pitch coke graphites showed an apparent decrease in their oxidation rate, more so than the petroleum coke graphites. These effects of the coke type reduced or nearly disappeared with an increasing temperature. The average activation energy determined for 608-808 deg. C was 161.5 ± 7.3 kJ/mol, showing that the dominant oxidation reaction occurred by a chemical control. A relationship between the oxidation rate and DOG was not observed

  10. Isotropic nuclear graphites; the effect of neutron irradiation

    International Nuclear Information System (INIS)

    Lore, J.; Buscaillon, A.; Mottet, P.; Micaud, G.

    1977-01-01

    Several isotropic graphites have been manufactured using different forming processes and fillers such as needle coke, regular coke, or pitch coke. Their properties are described in this paper. Specimens of these products have been irradiated in the fast reactor Rapsodie between 400 to 1400 0 C, at fluences up to 1,7.10 21 n.cm -2 PHI.FG. The results show an isotropic behavior under neutron irradiation, but the induced dimensional changes are higher than those of isotropic coke graphites although they are lower than those of conventional extruded graphites made with the same coke

  11. Pyrolysis and its potential use in nuclear graphite disposal

    International Nuclear Information System (INIS)

    Mason, J.B.; Bradbury, D.

    2001-01-01

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

  12. 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

  13. Microstructural characterization and pore structure analysis of nuclear graphite

    International Nuclear Information System (INIS)

    Kane, J.; Karthik, C.; Butt, D.P.; Windes, W.E.; Ubic, R.

    2011-01-01

    Graphite will be used as a structural and moderator material in next-generation nuclear reactors. While the overall nature of the production of nuclear graphite is well understood, the historic nuclear grades of graphite are no longer available. This paper reports the virgin microstructural characteristics of filler particles and macro-scale porosity in virgin nuclear graphite grades of interest to the Next Generation Nuclear Plant program. Optical microscopy was used to characterize filler particle size and shape as well as the arrangement of shrinkage cracks. Computer aided image analysis was applied to optical images to quantitatively determine the variation of pore structure, area, eccentricity, and orientation within and between grades. The overall porosity ranged between ∼14% and 21%. A few large pores constitute the majority of the overall porosity. The distribution of pore area in all grades was roughly logarithmic in nature. The average pore was best fit by an ellipse with aspect ratio of ∼2. An estimated 0.6-0.9% of observed porosity was attributed to shrinkage cracks in the filler particles. Finally, a preferred orientation of the porosity was observed in all grades.

  14. Statistical Comparison of the Baseline Mechanical Properties of NBG-18 and PCEA Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Mark C. Carroll; David T. Rohrbaugh

    2013-08-01

    High-purity graphite is the core structural material of choice in the Very High Temperature Reactor (VHTR), a graphite-moderated, helium-cooled design that is capable of producing process heat for power generation and for industrial process that require temperatures higher than the outlet temperatures of present nuclear reactors. The Baseline Graphite Characterization Program is endeavoring to minimize the conservative estimates of as-manufactured mechanical and physical properties by providing comprehensive data that captures the level of variation in measured values. In addition to providing a comprehensive comparison between these values in different nuclear grades, the program is also carefully tracking individual specimen source, position, and orientation information in order to provide comparisons and variations between different lots, different billets, and different positions from within a single billet. This report is a preliminary comparison between the two grades of graphite that were initially favored in the two main VHTR designs. NBG-18, a medium-grain pitch coke graphite from SGL formed via vibration molding, was the favored structural material in the pebble-bed configuration, while PCEA, a smaller grain, petroleum coke, extruded graphite from GrafTech was favored for the prismatic configuration. An analysis of the comparison between these two grades will include not only the differences in fundamental and statistically-significant individual strength levels, but also the differences in variability in properties within each of the grades that will ultimately provide the basis for the prediction of in-service performance. The comparative performance of the different types of nuclear grade graphites will continue to evolve as thousands more specimens are fully characterized from the numerous grades of graphite being evaluated.

  15. 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)

  16. 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.

  17. Statistical modeling of static strengths of nuclear graphites with relevance to structural design

    International Nuclear Information System (INIS)

    Arai, Taketoshi

    1992-02-01

    Use of graphite materials for structural members poses a problem as to how to take into account of statistical properties of static strength, especially tensile fracture stresses, in component structural design. The present study concerns comprehensive examinations on statistical data base and modelings on nuclear graphites. First, the report provides individual samples and their analyses on strengths of IG-110 and PGX graphites for HTTR components. Those statistical characteristics on other HTGR graphites are also exemplified from the literature. Most of statistical distributions of individual samples are found to be approximately normal. The goodness of fit to normal distributions is more satisfactory with larger sample sizes. Molded and extruded graphites, however, possess a variety of statistical properties depending of samples from different with-in-log locations and/or different orientations. Second, the previous statistical models including the Weibull theory are assessed from the viewpoint of applicability to design procedures. This leads to a conclusion that the Weibull theory and its modified ones are satisfactory only for limited parts of tensile fracture behavior. They are not consistent for whole observations. Only normal statistics are justifiable as practical approaches to discuss specified minimum ultimate strengths as statistical confidence limits for individual samples. Third, the assessment of various statistical models emphasizes the need to develop advanced analytical ones which should involve modeling of microstructural features of actual graphite materials. Improvements of other structural design methodologies are also presented. (author)

  18. Status of IAEA international data base on irradiated graphite properties with respect to HTR engineering issues

    International Nuclear Information System (INIS)

    Hacker, P.J.; Haag, G.

    2002-01-01

    The International Database on Irradiated Nuclear Graphite Properties contains data on the physical, chemical, mechanical and other relevant properties of graphites. Its purpose is to provide a platform that makes these properties accessible to approved users in the fields of nuclear power, nuclear safety and other nuclear science and technology applications. The database is constructed using Microsoft Access 97 software and has a controlled distribution by CD ROM to approved users. This paper describes the organisation and management of the database through administrative arrangements approved by the IAEA. It also outlines the operation of the database. The paper concludes with some remarks upon and illustrations of the usefulness of the database for the design and operation of HTR. (authors)

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

    Science.gov (United States)

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

    2017-02-01

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

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

    International Nuclear Information System (INIS)

    Mohd Reusmaazran Yusof; Yusof Abdullah

    2012-01-01

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

  1. The electrochemical properties of graphite and carbon

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  2. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-10-08

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

  4. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-11-15

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

  6. Nano-cracks in a synthetic graphite composite for nuclear applications

    Science.gov (United States)

    Liu, Dong; Cherns, David

    2018-05-01

    Mrozowski nano-cracks in nuclear graphite were studied by transmission electron microscopy and selected area diffraction. The material consisted of single crystal platelets typically 1-2 nm thick and stacked with large relative rotations around the c-axis; individual platelets had both hexagonal and cubic stacking order. The lattice spacing of the (0002) planes was about 3% larger at the platelet boundaries which were the source of a high fraction of the nano-cracks. Tilting experiments demonstrated that these cracks were empty, and not, as often suggested, filled by amorphous material. In addition to conventional Mrozowski cracks, a new type of nano-crack is reported, which originates from the termination of a graphite platelet due to crystallographic requirements. Both types are crucial to understanding the evolution of macro-scale graphite properties with neutron irradiation.

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  8. Fatigue properties of ductile cast iron containing chunky graphite

    Energy Technology Data Exchange (ETDEWEB)

    Ferro, P., E-mail: ferro@gest.unipd.it [Department of Management and Engineering, University of Padova, Stradella S. Nicola 3, I-36100 Vicenza (Italy); Lazzarin, P.; Berto, F. [Department of Management and Engineering, University of Padova, Stradella S. Nicola 3, I-36100 Vicenza (Italy)

    2012-09-30

    Highlights: Black-Right-Pointing-Pointer Experimental determination of high cycle fatigue properties of EN-GJS-400. Black-Right-Pointing-Pointer Evaluation of the influence of chunky graphite morphology on fatigue life. Black-Right-Pointing-Pointer Metallurgical analysis and microstructural parameters determination. Black-Right-Pointing-Pointer Nodule counting and nodularity rating. - Abstract: This work deals with experimental determination of high cycle fatigue properties of EN-GJS-400 ductile cast iron containing chunky graphite. Constant amplitude axial tests were performed at room temperature under a nominal load ratio R = 0. In order to evaluate the influence of chunky graphite morphology on fatigue life, fatigue tests were carried out also on a second set of specimens without this microstructural defect. All samples were taken from the core of a large casting component. Metallurgical analyses were performed on all the samples and some important microstructural parameters (nodule count and nodularity rating, among others) were measured and compared. It was found that a mean content of 40% of chunky graphite in the microstructure (with respect to total graphite content) does not influence significantly the fatigue strength properties of the analysed cast iron. Such result was attributed to the presence of microporosity detected on the surface fracture of the specimens by means of electron scanning microscope.

  9. Fatigue properties of ductile cast iron containing chunky graphite

    International Nuclear Information System (INIS)

    Ferro, P.; Lazzarin, P.; Berto, F.

    2012-01-01

    Highlights: ► Experimental determination of high cycle fatigue properties of EN-GJS-400. ► Evaluation of the influence of chunky graphite morphology on fatigue life. ► Metallurgical analysis and microstructural parameters determination. ► Nodule counting and nodularity rating. - Abstract: This work deals with experimental determination of high cycle fatigue properties of EN-GJS-400 ductile cast iron containing chunky graphite. Constant amplitude axial tests were performed at room temperature under a nominal load ratio R = 0. In order to evaluate the influence of chunky graphite morphology on fatigue life, fatigue tests were carried out also on a second set of specimens without this microstructural defect. All samples were taken from the core of a large casting component. Metallurgical analyses were performed on all the samples and some important microstructural parameters (nodule count and nodularity rating, among others) were measured and compared. It was found that a mean content of 40% of chunky graphite in the microstructure (with respect to total graphite content) does not influence significantly the fatigue strength properties of the analysed cast iron. Such result was attributed to the presence of microporosity detected on the surface fracture of the specimens by means of electron scanning microscope.

  10. A microstructurally based fracture model for nuclear graphite

    International Nuclear Information System (INIS)

    Burchell, T.D.

    1991-01-01

    This paper reports the physical basis of, and assumptions behind, a fracture model for nuclear graphites. Microstructurally related inputs, such as filler particle size, filler particle fracture toughness (K Ic ), density, pore size distribution, number of pores and specimen geometry (size and volume), are utilized in the model. The model has been applied to two graphites, Great Lakes Carbon Corporation grade H-451 and Toyo Tanso grade IG-110. For each graphite, the predicted tensile failure probabilities are compared with experimental data generated using ASTM Standard C-749 tensile test specimens. The predicted failure probabilities are in close agreement with the experimental data, particularly in the case of the H-451. The model is also shown to qualitatively predict the influence on the failure probabilities of changes in filler particle size, density, pore size, pore size distribution, number of pores and specimen geometry (stressed volume). The good performance is attributed to the sound physical basis of the model, which recognizes the dominant role of porosity in controlling crack initiation and propagation during graphite fracture. 8 refs., 12 figs., 1 tab

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-15

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

  12. 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

  13. 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

  14. 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

  15. High-temperature properties of commercial pyrolytic graphite

    International Nuclear Information System (INIS)

    Goetzel, C.G.

    1979-01-01

    Pyrolytic graphite is produced commercially as free-standing massive structures of continuously-nucleated and surface-nucleated varieties. Both are used for rocket nozzle components. The major design-significant thermal and mechanical properties of both material types are presented as a function of temperature up to 3000 K. They are compared with each other and with previous data. (author)

  16. Destruction of nuclear graphite using closed chamber incineration

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  17. Physical properties of C60 intercalated graphite films

    International Nuclear Information System (INIS)

    Nakahara, T; Hosomi, N; Taniguchi, J; Suzuki, M; Sato, T; Abe, K; Kuwahara, D; Ishikawa, M; Kato, M; Miura, K

    2007-01-01

    Recently, Miura and Tsuda have synthesized C 60 intercalated graphite film (C 60 /Gr) and reported that the C 60 /Gr consists of alternating close-packed C 60 monolayers and graphite layers. They also found that its frictional force is minimal up to the loading force of 100 nN using AFM [Miura K and Tsuda D 2005 e-J. Surf. Sci. Nanotech. 3 21] Thus, we have started to study the physical properties of C 60 /Gr and carried out NMR, Raman scattering and specific heat measurements. These results suggest that C 60 in C 60 /Gr rotates at room temperature

  18. Properties of screen-printed modified graphite layers

    Directory of Open Access Journals (Sweden)

    J. Walter

    2010-07-01

    Full Text Available During last years protection of the environment is one of the important problems that should be solved by modern technology. Theimportant problems are toxic gases emitted by conventional power plants. One of the methods that contribute to decreasing air pollution is manufacturing of cheap solar energy devices that could be applied in households. Among different type of fabrication technology of solar cells, DSSC technology looks like one of the interesting because it is relatively simple and low cost technology. Nowadays a lot of researcher groups making investigations to improve its setup, to get the cost reduction. The methods to achieve this goal were proposed in ISE (Germany as a concept of monolithic dye sensitised solar cell. One of the ideas of this solar cells setup is replacing expensive TCO electrode by much cheaper graphite electrode. Replacing TCO glass by graphite layer has to be done only in case of comparable properties of those both electrodes. There are some tested ideas of manufacturing that electrode and some of them are successfully applied. Presented work has been focused on preparation graphite conductive electrode for DSSC technology application, fabricated by screen–printing technique. Investigations concern new graphite past composition suitable for graphite layer preparation. It was been found that applying additive of titanium organic compound (Tyzor GBA to the past composition result in good properties, characterised by low resistance and good adhesion between graphite particles in the printed layer. Some tested layers prepared from proposed paste compositions characterised by better conductivity then applied in conventional DSSC cells counter electrode. The optimal addition of the modifier has not fixed yet.Among tested pastes the most promising results has been achieved for paste contained the biggest amount of Tyzor GBA.

  19. 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

  20. Mesostructure of graphite composite and its lifetime

    OpenAIRE

    Zhmurikov, Evgenij

    2015-01-01

    This review is devoted to the application of graphite and graphite composites in science and technology. Structure and electrical properties, as so technological aspects of producing of high strength artificial graphite and dynamics of its destruction are considered. These type of graphite are traditionally used in the nuclear industry. Generally, the review relies, on the original results and concentrates on actual problems of application and testing of graphite materials in modern nuclear p...

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

    International Nuclear Information System (INIS)

    2001-05-01

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

  2. Oxidation parameters of nuclear graphite for HTGR air-ingress

    International Nuclear Information System (INIS)

    Kim, E.S.; No, H.C.

    2004-01-01

    In order to investigate chemical behaviors of the graphite during an air-ingress accident in HTGR, the kinetic tests on nuclear graphite IG-110 were performed in chemical reaction dominant regime. In the present experiment, inlet gas flow rate ranged between 8 and 18 SLPM, graphite temperatures and oxygen mole fraction ranged from 540 to 630degC and from 3 to 30% respectively. The test section was made of a quartz tube having 75 mm diameter and 750 mm length and the test specimen machined to the size of 21 mm diameter and 30 mm length was supported at the center of it by the alumina rod. The 15 kW induction heater was installed around the outside of test section to heat the specimen and its temperature was measured by 2 infrared thermometers. The oxidation rate was calculated from the gas concentration analysis between inlet and outlet using NDIR (non-dispersive infrared) gas analyzer. As a result the activation energy (Ea) and the order of reaction (n) were determined within 95% confidence level and the qualitative characteristics of the two parameters were also widely investigated by experimental and analytical methods. (author)

  3. 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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  5. 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

  6. Property changes in graphite irradiated at changing irradiation temperature

    International Nuclear Information System (INIS)

    Price, R.J.; Haag, G.

    1979-07-01

    Design data for irradiated graphite are usually presented as families of isothermal curves showing the change in physical property as a function of fast neutron fluence. In this report, procedures for combining isothermal curves to predict behavior under changing irradiation temperatures are compared with experimental data on irradiation-induced changes in dimensions, Young's modulus, thermal conductivity, and thermal expansivity. The suggested procedure fits the data quite well and is physically realistic

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

    International Nuclear Information System (INIS)

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

    2001-01-01

    Over the last twenty years, numerous proposals have been made for the long-term treatment of radioactive graphite waste. These plans have ranged from sea dumping through incineration to land-based disposal, sometimes preceded by a variable period of 'safe-storage' within the original reactor containment, to allow for the decay of shorter-lived isotopes ahead of dismantling. A number of novel chemical or physical pre-treatments of the graphite, with the objective of facilitating its subsequent disposal or improving the environmental consequences of the chosen disposal route, have also been suggested. There are patents issued on systems for transmutation of long-lived isotopes to reduce the radiological consequences of disposal of intact graphite, and for separation of certain isotopes such as carbon-14 from the matrix in an incineration process. Although these far-reaching proposals are not apparently cost-effective, scope for cost-recovery does exist, i.e., in terms of disposal of the separated carbon-14 in cements used for immobilisation of other radioactive solid waste materials. More recently, political and environmental factors have further complicated the issue. Nuclear regulators are challenging the proposed length of 'safe-storage' schemes on the basis that essential knowledge on the reactor materials may be lost in the interim. International agreements such as OSPAR have effectively eliminated the possibility for disposal at sea, whilst public opinion is strongly expressed against any expansion of existing land-based disposal sites or the creation of new ones. As a particular example, the United Kingdom authorities recently denied to the official body charged with the development of a deep repository the necessary planning consents to develop an exploratory rock-structure laboratory on the most favoured site. The current drive towards minimising or eliminating any radioactivity release to the environment has the unintended consequence of causing the waste

  8. Observations in the statistical analysis of NBG-18 nuclear graphite strength tests

    International Nuclear Information System (INIS)

    Hindley, Michael P.; Mitchell, Mark N.; Blaine, Deborah C.; Groenwold, Albert A.

    2012-01-01

    Highlights: ► Statistical analysis of NBG-18 nuclear graphite strength test. ► A Weibull distribution and normal distribution is tested for all data. ► A Bimodal distribution in the CS data is confirmed. ► The CS data set has the lowest variance. ► A Combined data set is formed and has Weibull distribution. - Abstract: The purpose of this paper is to report on the selection of a statistical distribution chosen to represent the experimental material strength of NBG-18 nuclear graphite. Three large sets of samples were tested during the material characterisation of the Pebble Bed Modular Reactor and Core Structure Ceramics materials. These sets of samples are tensile strength, flexural strength and compressive strength (CS) measurements. A relevant statistical fit is determined and the goodness of fit is also evaluated for each data set. The data sets are also normalised for ease of comparison, and combined into one representative data set. The validity of this approach is demonstrated. A second failure mode distribution is found on the CS test data. Identifying this failure mode supports the similar observations made in the past. The success of fitting the Weibull distribution through the normalised data sets allows us to improve the basis for the estimates of the variability. This could also imply that the variability on the graphite strength for the different strength measures is based on the same flaw distribution and thus a property of the material.

  9. 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.

  10. 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

  11. 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

  12. Direct reading spectrochemical analysis of nuclear graphite; Analisis espectroquimico de lectura directa de grafito nuclear

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

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

  13. Fracture behavior of nuclear graphites under tensile impact loading

    International Nuclear Information System (INIS)

    Ugachi, Hirokazu; Ishiyama, Shintaro; Eto, Motokuni

    1994-01-01

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

  14. Balanced improvement of high performance concrete material properties with modified graphite nanomaterials

    Science.gov (United States)

    Peyvandi, Amirpasha

    Graphite nanomaterials offer distinct features for effective reinforcement of cementitious matrices in the pre-crack and post-crack ranges of behavior. Thoroughly dispersed and well-bonded nanomaterials provide for effective control of the size and propagation of defects (microcracks) in matrix, and also act as closely spaced barriers against diffusion of moisture and aggressive solutions into concrete. Modified graphite nanomaterials can play multi-faceted roles towards enhancing the mechanical, physical and functional attributes of concrete materials. Graphite nanoplatelets (GP) and carbon nanofibers (CNF) were chosen for use in cementitious materials. Experimental results highlighted the balanced gains in diverse engineering properties of high-performance concrete realized by introduction of graphite nanomaterials. Nuclear Magnetic Resonance (NMR) spectroscopy was used in order to gain further insight into the effects of nanomaterials on the hydration process and structure of cement hydrates. NMR exploits the magnetic properties of certain atomic nuclei, and the sensitivity of these properties to local environments to generate data which enables determination of the internal structure, reaction state, and chemical environment of molecules and bulk materials. 27 Al and 29Si NMR spectroscopy techniques were employed in order to evaluate the effects of graphite nanoplatelets on the structure of cement hydrates, and their resistance to alkali-silica reaction (ASR), chloride ion diffusion, and sulfate attack. Results of 29Si NMR spectroscopy indicated that the percent condensation of C-S-H in cementitious paste was lowered in the presence of nanoplatelets at the same age. The extent of chloride diffusion was assessed indirectly by detecting Friedel's salt as a reaction product of chloride ions with aluminum-bearing cement hydrates. Graphite nanoplatelets were found to significantly reduce the concentration of Friedel's salt at different depths after various periods

  15. Kinetics of Chronic Oxidation of NBG-17 Nuclear Graphite by Water Vapor

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-01

    This report presents the results of kinetic measurements during accelerated oxidation tests of NBG-17 nuclear graphite by low concentration of water vapor and hydrogen in ultra-high purity helium. The objective is to determine the parameters in the Langmuir-Hinshelwood (L-H) equation describing the oxidation kinetics of nuclear graphite in the helium coolant of high temperature gas-cooled reactors (HTGR). Although the helium coolant chemistry is strictly controlled during normal operating conditions, trace amounts of moisture (predictably < 0.2 ppm) cannot be avoided. Prolonged exposure of graphite components to water vapor at high temperature will cause very slow (chronic) oxidation over the lifetime of graphite components. This behavior must be understood and predicted for the design and safe operation of gas-cooled nuclear reactors. The results reported here show that, in general, oxidation by water of graphite NBG-17 obeys the L-H mechanism, previously documented for other graphite grades. However, the characteristic kinetic parameters that best describe oxidation rates measured for graphite NBG-17 are different than those reported previously for grades H-451 (General Atomics, 1978) and PCEA (ORNL, 2013). In some specific conditions, certain deviations from the generally accepted L-H model were observed for graphite NBG-17. This graphite is manufactured in Germany by SGL Carbon Group and is a possible candidate for the fuel elements and reflector blocks of HTGR.

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  17. Electrochemical tuning of optical properties of graphitic quantum dots

    International Nuclear Information System (INIS)

    Ge, Juan; Li, Yan; Zhang, Bo-Ping; Ma, Ning; Wang, Jun; Pu, Chang; Xiang, Ying-Chang

    2015-01-01

    Graphitic quantum dots (GQDs), as a new class of quantum dots, possess unique properties. Among the various reported approaches for their fabrication, electrochemical method possesses numerous advantages compared with others. In particular, the formation process of the GQDs could be precisely controlled by this method through adjusting the electrochemical parameters and environment. In this study, GQDs with multi-color fluorescence (FL) were obtained by this method through tuning only the applied potential window of cycling voltammetry. The luminescence mechanism of those GQDs was discussed and explained by the ultraviolet (UV)–visible, photoluminescence (PL), and photoluminescence excitation (PLE) spectra. The influence of the applied potential window on the PL properties of GQDs and the relationship between the degree of surface oxidation and PL properties were also investigated. - Highlights: • We produced the graphite quantum dots (GQDs) by an electrochemical method. • We changed the applied potentials of cycling voltammetry (CV). • Varying of applied potentials changed surface oxygen-containing groups of GQDs. • Higher surface oxidation degree resulted in the red-shift of PL spectra

  18. Examination of Experimental Data for Irradiation - Creep in Nuclear Graphite

    Science.gov (United States)

    Mobasheran, Amir Sassan

    The objective of this dissertation was to establish credibility and confidence levels of the observed behavior of nuclear graphite in neutron irradiation environment. Available experimental data associated with the OC-series irradiation -induced creep experiments performed at the Oak Ridge National Laboratory (ORNL) were examined. Pre- and postirradiation measurement data were studied considering "linear" and "nonlinear" creep models. The nonlinear creep model considers the creep coefficient to vary with neutron fluence due to the densification of graphite with neutron irradiation. Within the range of neutron fluence involved (up to 0.53 times 10^{26} neutrons/m ^2, E > 50 KeV), both models were capable of explaining about 96% and 80% of the variation of the irradiation-induced creep strain with neutron fluence at temperatures of 600^circC and 900^circC, respectively. Temperature and reactor power data were analyzed to determine the best estimates for the actual irradiation temperatures. It was determined according to thermocouple readouts that the best estimate values for the irradiation temperatures were well within +/-10 ^circC of the design temperatures of 600^circC and 900 ^circC. The dependence of the secondary creep coefficients (for both linear and nonlinear models) on irradiation temperature was determined assuming that the variation of creep coefficient with temperature, in the temperature range studied, is reasonably linear. It was concluded that the variability in estimate of the creep coefficients is definitely not the results of temperature fluctuations in the experiment. The coefficients for the constitutive equation describing the overall growth of grade H-451 graphite were also studied. It was revealed that the modulus of elasticity and the shear modulus are not affected by creep and that the electrical resistivity is slightly (less than 5%) changed by creep. However, the coefficient of thermal expansion does change with creep. The consistency of

  19. 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)

  20. 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)

  1. Methods for measuring nuclear properties of materials, Safety coefficient method and measurement of effective absorption coefficient of graphite by safety coefficient method; Razvijanje metoda merenja nuklearnih karakteristika materijala, Razrada metode koeficijenta opasnosti i merenje efektivnog apsorpcionog preseka grafita metodom koeficijenta opasnosti

    Energy Technology Data Exchange (ETDEWEB)

    Maglic, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-11-15

    Reactivity of a reactor depends on production, absorption and leaking of neutrons. Change of absorption causes reactivity change, and this fact is used for determining the neutron absorption cross section for the sample inserted in the reactor core. Method for determining the absorption cross section based on reactivity change is called method of safety coefficient. Measurements of neutron absorption cross section for graphite was done in the RA reactor vertical experimental channel VK-5. taking into account the results obtained for five types of graphite this method is considered to be reliable for use. Comparison of nuclear properties of different types of graphite was done as well. Reaktivnost reaktora zavisi od proizvodnje neutrona, apsorpcije i isticanja neutrona. Promena apsorpcije izaziva promene reaktivnosti reaktora pa se ova osobina koristi za odedjivanje neutronskog apsorpcionog preseka uzorka koji se unosi u reaktor. Metoda merenja apsorpcionog preseka na bazi promene reaktivnosti nazvana je metodom koeficijenta opasnosti. Merenje apsorpcionog preseka grafita uradjeno je na reaktoru RA u vertikalnom eksperimentalnom kanalu VK-5. S obzirom na rezultate koji su dobijeni za pet vrsta grafita moze se smatrati da je opravdano koriscenje ove metode. Izvrseno je i poredjenje nuklearnih osobina pomenutih tipova grafita.

  2. 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)

  3. A discussion of possible mechanisms affecting fission product transport in irradiated and unirradiated nuclear grade graphite

    International Nuclear Information System (INIS)

    Firth, M.J.

    1977-09-01

    137 Cs, 85 Sr, and sup(110m)Ag adsorption experiments were conducted on three graphite powders with differing amounts of specific basal and edge surface areas. No direct proportionality was found between the specific amounts of the isotopes adsorbed and either of the surface characteristics. There appears to be some correlation with the specific basal surface area despite the fact that each isotope behaves differently. Factors that might influence the adsorption behaviour of Cs and Ag during reactor irradiation and heat treatment of nuclear grade graphites are discussed. These include the form of Cs with the graphite surface. A model based on Cs adsorption at vacancy clusters is used to analyse adsorption experiments. A possible explanation for the behaviour of Ag through the migration of graphite impurities from the bulk of the graphite to the pore surface is also discussed. (author)

  4. The reaction of unirradiated and irradiated nuclear graphites with water vapor in helium

    International Nuclear Information System (INIS)

    Imai, Hisashi; Nomura, Shinzo; Kurosawa, Takeshi; Fujii, Kimio; Sasaki, Yasuichi

    1980-10-01

    Nuclear graphites more than 10 brands were oxidized with water vapor in helium and then some selected graphites were irradiated with fast neutron in the Japan Materials Testing Reactor to clarify the effect of radiation damage of graphite on their reaction behaviors. The reaction was carried out under a well defined condition in the temperature range 800 -- 1000 0 C at concentrations of water vapor 0.38 -- 1.30 volume percent in helium flow of total pressure of 1 atm. The chemical reactivity of graphite irradiated at 1000 +- 50 0 C increased linearly with neutron fluence until irradiation of 3.2 x 10 21 n/cm 2 . The activation energy for the reaction was found to decrease with neutron fluence for almost all the graphites, except for a few ones. The order of reaction increased from 0.5 for the unirradiated graphite to 1.0 for the graphite irradiated up to 6.0 x 10 20 n/cm 2 . Experiment was also performed to study a superposed effect between the influence of radiation damage of graphite and the catalytic action of barium on the reaction rate, as well as the effect of catalyser of barium. It was shown that these effects were not superposed upon each other, although barium had a strong catalytic action on the reaction. (author)

  5. Thermal properties of graphite oxide, thermally reduced graphene and chemically reduced graphene

    Science.gov (United States)

    Jankovský, Ondřej; Sedmidubský, David; Lojka, Michal; Sofer, Zdeněk

    2017-07-01

    We compared thermal behavior and other properties of graphite oxide, thermally reduced graphene and chemically reduced graphene. Graphite was oxidized according to the Hofmann method using potassium chlorate as oxidizing agent in strongly acidic environment. In the next step, the formed graphite oxide was chemically or thermally reduced yielding graphene. The mechanism of thermal reduction was studied using STA-MS. Graphite oxide and both thermally and chemically reduced graphenes were analysed by SEM, EDS, elemental combustion analysis, XPS, Raman spectroscopy, XRD and BET. These findings will help for the large scale production of graphene with appropriate chemical composition.

  6. Nuclear graphite based on coal tar pitch; behavior under neutron irradiation between 400 and 14000C

    International Nuclear Information System (INIS)

    Mottet, P.; Fillatre, A.; Schill, R.; Micaud, G.

    1977-01-01

    Two nuclear grades of coal tar pitch coke graphites have been developed and tested under neutron irradiation. The neutron irradiation induced dimensional changes between 400 and 1400 0 C, at fluences up to 1,2.10 22 n.cm -2 PHI.FG show a behavior comparable to anisotropic petroleum coke graphites. Less than 10% variation in thermal expansion, maximum decrease by a factor four in thermal conductivity, and large increase of the Young modulus have been observed

  7. Recent Advances in Preparation, Structure, Properties and Applications of Graphite Oxide.

    Science.gov (United States)

    Srivastava, Suneel Kumar; Pionteck, Jürgen

    2015-03-01

    Graphite oxide, also referred as graphitic oxide or graphitic acid, is an oxidized bulk product of graphite with a variable composition. However, it did not receive immense attention until it was identified as an important and easily obtainable precursor for the preparation of graphene. This inspired many researchers to explore facts related to graphite oxide in exploiting its fascinating features. The present article culminates up-dated review on different preparative methods, morphology and characterization of physical/chemical properties of graphite oxide by XRD, XPS, FTIR, Raman, NMR, UV-visible, and DRIFT analyses. Finally, recent developments on intercalation and applications of GO in multifaceted areas of catalysis, sensor, supercapacitors, water purification, hydrogen storage and magnetic shielding etc. has also been reviewed.

  8. Understanding of mechanical properties of graphite on the basis of mesoscopic microstructure (review)

    International Nuclear Information System (INIS)

    Ishihara, M.; Shibata, T.; Takahashi, T.; Baba, S.; Hoshiya, T.

    2002-01-01

    With the aim of nuclear application of ceramics in the high-temperature engineering field, the authors have investigated the mesoscopic microstructure related to the mechanical and thermal properties of ceramics. In this paper, recent activities concerning mechanical properties, strength and Young's modulus are presented. In the strength research field, the brittle fracture model considering pore/grain mesoscopic microstructure was expanded so as to render possible an estimation of the strength under stress gradient conditions. Furthermore, the model was expanded to treat the pore/crack interaction effect. The performance of the developed model was investigated from a comparison with experimental data and the Weibull strength theory. In the field of Young's modulus research, ultrasonic wave propagation was investigated using the pore/wave interaction model. Three kinds of interaction modes are treated in the model. The model was applied to the graphite, and its applicability was investigated through comparison with experimental data. (authors)

  9. Structural and Kinetic Properties of Graphite Intercalation Compounds

    Science.gov (United States)

    1983-04-29

    The exfoliation of graphite-FeCl 3NH has been used for making blankets for the extinction of metal fires [12). In addition. exfoliated graphite is...FeCl3-oH3 has been used (Aerotech GCma, 0.5 MHz wideband) equipped with for making blankets for the extinction of metal fires (3). In addition

  10. Nuclear Magnetic Resonance Study of Fluorine-Graphite Intercalation Compounds

    International Nuclear Information System (INIS)

    Panich, A.M.; Goren, S.D.; Nakajima, T.; Vieth, H.-M.; Privalov, A.

    1998-01-01

    To study the origin of semimetal-metal and metal-insulator transformations, localization effects and C-E bonding in fluorine-intercalated graphite C x F, 13 C and 19 F NMR investigations have been carried out for a wide range of fluorine content, 3.8 8, are attributed to mobile fluorine acceptor species which are responsible for the increase of electric conductivity in the dilute compound. When increasing the fluorine content to x ∼ 8 corresponding to the maximum electric conductivity, covalent C-P bonds start to oc- cur. The number of these bonds grows with fluorine content resulting in the decrease in conductivity which is caused by a percolation mechanism rather than by a change in bond length. A difference in 19 F chemical shift for fluorine-intercalated graphite C x F and covalent graphite fluoride (CF) n has been observed and is attributed to different C-P bonding in these compounds

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-11

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  14. Application of a micromechanics model to the overall properties of heterogeneous graphite

    International Nuclear Information System (INIS)

    Berre, C.; Mummery, P.M.; Marsden, B.J.; Mori, T.; Withers, P.J.

    2008-01-01

    This paper deals with the overall properties of polycrystalline graphite, a material mainly composed of voids and dense inhomogeneities embedded in a less dense matrix. First, we examine the overall average elastic properties and conductivities of such a material. Second, we evaluate the void shape effects on the overall Young's modulus. Finally, we compare the results obtained from the analytical model with experimental data from radiolytic oxidation of graphite

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

    Science.gov (United States)

    Pierce, Robert A.

    2017-10-17

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

  16. Industrial nuclear property

    International Nuclear Information System (INIS)

    Lepetre, M.

    1976-01-01

    The first requests for patents for the use of nuclear power filed in France in 1939. This paper reviews the regulations on industrial nuclear property in various countries. The patenting system in several socialist countries is characterized by the fact that inventions on the production and use of radioactive materials may not be patented. This equally applies in India. In the United States, this type of invention may be patented except for those involving military uses and which must be notified to the federal authorities. In France, all industrial nuclear property is grouped under the same body, Brevatome, created in 1958, which enables the allocation of rights to be negotiated between the different interested parties, the Atomic Energy Commission (CEA), Electricite de France (EDF) and private industry. Under the Euratom Treaty, all inventions, even those governed by secrecy in Member countries, must be communicated to the Commission of the European Communities. (NEA) [fr

  17. 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.

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  19. Eddy current testing on structures of nuclear-grade IG-110 graphite for acceptance test in HTTR

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Saikusa, Akio; Iyoku, Tatsuo

    1993-09-01

    Core and core support graphite structures in the HTTR are mainly made of IG-110 graphite which is fine-grained isotropic and nuclear-grade. Nondestructive inspection with eddy current testing is planned to be applied to these graphite structures. Eddy current testing is widely applied to metallic structures and its testing method has been already established. On the other hand, the characteristics of graphite are quite different in micro-structure from these of metals. Therefore, the eddy current testing method provided for metallic structures can not be applied directly to graphite structures. Thus the eddy current testing method and condition were established for the graphite structures made of IG-110 graphite. (author)

  20. Influence of graphite contamination on the optical properties of transparent spinel obtained by spark plasma sintering

    International Nuclear Information System (INIS)

    Bernard-Granger, G.; Benameur, N.; Guizard, C.; Nygren, M.

    2009-01-01

    The optical properties of transparent spinel sintered by spark plasma sintering have been investigated for incident electromagnetic radiations with wavelengths in the range 0.2-2 μm. It is shown that residual porosities and second-phase graphite particles have a strong influence on the in-line transmittance. Because of the graphite particles, the in-line transmittance measured does not approach that of monocrystalline spinel for wavelengths above 1 μm

  1. Synthesis, characterization and magnetic properties of room-temperature nanofluid ferromagnetic graphite

    OpenAIRE

    Souza, N. S.; Sergeenkov, S.; Speglich, C.; Rivera, V. A. G.; Cardoso, C. A.; Pardo, H.; Mombru, A. W.; Rodrigues, A. D.; de Lima, O. F.; Araujo-Moreira, F. M.

    2009-01-01

    We report the chemical synthesis route, structural characterization, and physical properties of nanofluid magnetic graphite (NFMG) obtained from the previously synthesized bulk organic magnetic graphite (MG) by stabilizing the aqueous ferrofluid suspension with an addition of active cationic surfactant. The measured magnetization-field hysteresis curves along with the temperature dependence of magnetization confirmed room-temperature ferromagnetism in both MG and NFMG samples. (C) 2009 Americ...

  2. Influence of expanded graphite (EG) and graphene oxide (GO) on physical properties of PET based nanocomposites

    OpenAIRE

    Paszkiewicz Sandra; Nachman Małgorzata; Szymczyk Anna; Špitalský Zdeno; Mosnáček Jaroslav; Rosłaniec Zbigniew

    2014-01-01

    This work is the continuation and refinement of already published communications based on PET/EG nanocomposites prepared by in situ polymerization1, 2. In this study, nanocomposites based on poly(ethylene terephthalate) with expanded graphite were compared to those with functionalized graphite sheets (GO). The results suggest that the degree of dispersion of nanoparticles in the PET matrix has important effect on the structure and physical properties of the nanocomposites. The existence of gr...

  3. Effects of the temperature and the irradiation on the behaviour of chlorine 37 in nuclear graphite: consequences on the mobility of chlorine 36 in irradiated graphites

    International Nuclear Information System (INIS)

    Blondel, Antoine

    2013-01-01

    This thesis deals with the studies of the management of irradiated graphite wastes issued from the dismantling of the UNGG French reactors. This work focuses on the behavior of 36 Cl. This radionuclide is mainly issued through the neutron activation of 35 Cl by the reaction 35 Cl(n, γ) 36 Cl, pristine chlorine being an impurity of nuclear graphite, present at the level of some at.ppm. 36 Cl is a long lived radionuclide (about 300,000 years) and is highly soluble in water and mobile in concrete and clay. The solubilization of 36 Cl is controlled by the water accessibility into irradiated graphite pores as well as by factors related to 36 Cl itself such as its chemical speciation and its location within the irradiated graphite. Both speciation and chlorine location should strongly influence its behaviour and need to be taken into account for the choice of liable management options. However, data on radioactive chlorine features are difficult to assess in irradiated graphite and are mainly related to detection sensitivity problems. In this context, we simulated and evaluated the impact of the temperature, the irradiation and the radiolytic oxidation on the chlorine 36 behaviour. In order to simulate the presence of 36 Cl, we implanted 37 Cl into virgin nuclear graphite. Ion implantation has been widely used to study the lattice location, the diffusion and the release of fission and activation products in nuclear materials. Our results on the comparative effects of the temperature and the irradiation show that chlorine occurs in irradiated graphite on temperature and electronic and nuclear irradiation improve this effect. (author)

  4. Average nuclear surface properties

    International Nuclear Information System (INIS)

    Groote, H. von.

    1979-01-01

    The definition of the nuclear surface energy is discussed for semi-infinite matter. This definition is extended also for the case that there is a neutron gas instead of vacuum on the one side of the plane surface. The calculations were performed with the Thomas-Fermi Model of Syler and Blanchard. The parameters of the interaction of this model were determined by a least squares fit to experimental masses. The quality of this fit is discussed with respect to nuclear masses and density distributions. The average surface properties were calculated for different particle asymmetry of the nucleon-matter ranging from symmetry beyond the neutron-drip line until the system no longer can maintain the surface boundary and becomes homogeneous. The results of the calculations are incorporated in the nuclear Droplet Model which then was fitted to experimental masses. (orig.)

  5. Microwave absorbing property of a hybrid absorbent with carbonyl irons coating on the graphite

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Yonggang, E-mail: xuyonggang221@163.com [Science and Technology on Electromagnetic Scattering Laboratory, Shanghai, 200438 (China); Yan, Zhenqiang; Zhang, Deyuan [Bionic and Micro/Nano/Bio Manufacturing Technology Research Center, School of Mechanical Engineering and Automation, Beihang University, Beijing 100191 (China)

    2015-11-30

    Graphical abstract: The absorbing property could be enhanced as the CIPs coated on the graphite. - Highlights: • Absorbers filled with CIPs coating on the graphite was fabricated. • The permittivity and permeability increased as CIPs coated. • The CIP materials enhanced the electromagnetic property. • The graphite coated CIPs were effective in 2–18 GHz. - Abstract: The hybrid absorbent filled with carbonyl iron particles (CIPs) coating on the graphite was prepared using a chemical vapor decomposition (CVD) process. X-ray diffraction (XRD) patterns were done to analyze the particle crystal grain structure. The complex permittivity and permeability were measured using a vector network analyzer in the frequency range of 2–18 GHz. The results showed that α-Fe appeared in the super-lattice diffraction peaks in XRD graph. The composites added CIPs coating on the graphite had a higher permittivity and imaginary permeability due to the superior microwave dielectric loss and magnetic loss of the CIPs. The reflection loss (RL) result showed that composites filled with 5 vol% Fe-graphite had an excellent absorbing property in the 2–18 GHz, the minimum RL was −25.14 dB at 6 mm and −26.52 dB at 8 mm, respectively.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  7. Effects of porosity and temperature on oxidation behavior in air of selected nuclear graphites

    International Nuclear Information System (INIS)

    Chen Dongyue; Li Zhengcao; Miao Wei; Zhang Zhengjun

    2012-01-01

    Nuclear graphite endures gas oxidation in High Temperature Gas-cooled Reactor (HTGR), which may threaten the safety of reactor. To study the oxidation behavior of nuclear graphite, weight loss curve is usually measured through Thermo Gravimetric Analysis (TGA) method. In this work, three brands of nuclear graphite for HTGR (i.e., HSM-SC, IG-11, and NBG-18) are oxidized under 873 and 1073 K in open air, and their weight loss curves are obtained. The acceleration of oxidizing rate is observed for both HSM-SC and IG-11, and is attributed to the large porosity increase during oxidation process. For HSM-SC, the porosity increase comes from preferential binder oxidation, and thus its binder quality shall be improved to obtain better oxidation resistance. Temperature effects on oxidation for HSM-SC are also studied, which shows that oxidizing gas tends to be exhausted at graphite surface at high temperature instead of penetrate into the interior of bulk. (author)

  8. 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.

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

    International Nuclear Information System (INIS)

    Hacker, P.J.

    2001-02-01

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

  10. Study of a nuclear graphite waste 14C decontamination process by CO2 gasification

    International Nuclear Information System (INIS)

    Pageot, Justin

    2014-01-01

    The decommissioning of French gas cooled nuclear reactors (UNGG), all arrested since 1994, will generate 23,000 tons of graphite waste classified Low Level and Long Lived and notably containing 14 C. The aim of this thesis is to study a new method for selective extraction of this radionuclide by CO 2 gasification.The multi-scale organization of virgin and irradiated graphite has been studied by a coupling between microspectrometry Raman and transmission electron microscopy. With the neutron fluence, the structure degrades and the nano-structure can be greatly changed. In extreme cases, the lamellar nano-structure nuclear graphite has become nano-porous. Furthermore, these damages are systematically heterogeneous. An orientation effect of 'crystallites', shown experimentally by ion implantation, could be a cause of these heterogeneities.This study also showed that from a specific fluence, there is an important development of nano-porous zones coinciding with a dramatic 14 C concentration increase. This radionuclide could be preferentially concentrated in the nano-porous areas which are potentially more reactive than the remaining laminar areas which could be less rich in 14 C. This process by CO 2 gasification was firstly tested on 'analogous' non-radioactive materials (mechanically milled graphite). These tests confirmed, for temperatures between 950 and 1000 C, the selective and complete elimination of nano-porous areas.Tests were then carried out on graphite waste from Saint-Laurent-des-Eaux A2 and G2 reactors. The results are promising with notably the quarter of 14 C inventory extracted for a weight loss of only few percent. Up to 68 % of 14 C inventory was extracted, but with an important gasification. Thus, this treatment could allow extracting selectively a share of 14 C inventory (mobile or linked to nano-porous areas) and allows imagining alternative scenarios for graphite waste managing. (author) [fr

  11. AGC-2 Graphite Preirradiation Data Package

    Energy Technology Data Exchange (ETDEWEB)

    David Swank; Joseph Lord; David Rohrbaugh; William Windes

    2012-10-01

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

  12. 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)

  13. Description of nuclear properties

    International Nuclear Information System (INIS)

    Faessler, A.

    1991-01-01

    The lectures want to give a survey about new developments in the description of nuclei. In a first chapter we try to derive nuclear properties from the basis theory of quantum chromodynamics. This is not rigorously possible. There are still many cracks in the bridge between QCD and nuclear structure. The basic ingredient for nuclear structure calculations is the nucleon-nucleon interaction. We shall discuss the nucleon-nucleon interaction in a quark model. In a further chapter we discuss the way to come from a bare nucleon-nucleon interaction in free space to an effective nucleon-nucleon interaction in a limited model space for nuclear structure calculations. Such nuclear structure calculations can be done as shell model calculations. But they are due to the large number of configurations limited to light nuclei. We discuss possibilities (MONSTER and VAMPIR) to enlarge the model space for medium heavy and heavy nuclei. As the example of the low lying isovector 1 + states we discuss collective models (Bohr - Mottelson Model, interacting Boson Model) with proton and neutron degrees of freedom. The same states can also be described microscopically with the Quasi-Particle Random Phase Approximation (QRPA). We discuss the removal of spurious states in RPA. We also discuss the calculation of form factors and compare with inelastic electron scattering data. Finally we apply QRPA to the double-beta decay. Grand unified models predict, that the neutrino is identical with his antiparticle, that it has a finite mass and a weak right-handed interaction. If these properties are found the standard model of the strong and the electro-weak interaction can not be correct. Presently we can only derive from lower limits of the half lives of neutrinoless double-beta decays upper limits of the neutrino mass and of the right-handedness of the weak interaction and lower limits of the mass of the right handed heavy vector boson, if a specific grand unified model is given. (author)

  14. Microstructure, elastic and inelastic properties of partially graphitized biomorphic carbons

    Science.gov (United States)

    Orlova, T. S.; Kardashev, B. K.; Smirnov, B. I.; Gutierrez-Pardo, A.; Ramirez-Rico, J.; Martinez-Fernandez, J.

    2015-03-01

    The microstructural characteristics and amplitude dependences of the Young's modulus E and internal friction (logarithmic decrement δ) of biocarbon matrices prepared by beech wood carbonization at temperatures T carb = 850-1600°C in the presence of a nickel-containing catalyst have been studied. Using X-ray diffraction and electron microscopy, it has been shown that the use of a nickel catalyst during carbonization results in a partial graphitization of biocarbons at T carb ≥ 1000°C: the graphite phase is formed as 50- to 100-nm globules at T carb = 1000°C and as 0.5- to 3.0-μm globules at T carb = 1600°C. It has been found that the measured dependences E( T carb) and δ( T carb) contain three characteristic ranges of variations in the Young's modulus and logarithmic decrement with a change in the carbonization temperature: E increases and δ decreases in the ranges T carb 1300°C; in the range 1000 biocarbons carbonized in the presence of nickel correlates with the evolution of their microstructure. The largest values of E are obtained for samples with T carb = 1000 and 1600°C. However, the samples with T carb = 1600°C exhibit a higher susceptibility to microplasticity due to the presence of a globular graphite phase that is significantly larger in size and total volume.

  15. Multi-scale characterization and modelling of damage evolution in nuclear Gilsocarbon graphite

    NARCIS (Netherlands)

    Liu, D.; Heard, P.; Savija, B.; Smith, G.; Schlangen, E.; Flewitt, P.

    2015-01-01

    In the present work, the microstructure and mechanical properties of Gilsocarbon graphite have been characterized over a range of length-scales. Optical imaging, combined with 3D X-ray computed tomography and 3D high-resolution tomography based on focus ion beam milling has been adopted for

  16. Modeling the overall heat conductive and convective properties of open-cell graphite foam

    International Nuclear Information System (INIS)

    Tee, C C; Yu, N; Li, H

    2008-01-01

    This work develops analytic models on the overall thermal conductivity, pressure drop and overall convective heat transfer coefficient of graphite foam. The models study the relationship between the overall heat conductive and convective properties, and foam microstructure, temperature, foam surface friction characteristics and cooling fluid properties. The predicted thermal conductivity, convective heat transfer coefficient and pressure drop agree well with experimental data

  17. Improved mechanical and functional properties of elastomer/graphite nanocomposites prepared by latex compounding

    International Nuclear Information System (INIS)

    Yang Jian; Tian Ming; Jia Qingxiu; Shi Junhong; Zhang Liqun; Lim Szuhui; Yu Zhongzhen; Mai Yiuwing

    2007-01-01

    The facile latex approach has been adopted to finely incorporate graphite nanosheets into elastomeric polymer matrix to obtain high-performance elastomeric nanocomposites with improved mechanical properties and functional properties. Scanning electron microscopy, transmission electron microscopy and X-ray diffraction experiments show that the nanostructures of the final nanocomposites exhibit a high degree of exfoliation and intercalation of graphite in the nitrile-butadiene rubber (NBR) matrix. Mechanical and dynamic-mechanical tests demonstrate that the NBR/graphite nanocomposites possess greatly increased elastic modulus and tensile strength, and desirably strong interfaces. The unexpected self-crosslinking of elastomer/graphite nanocomposites was discovered and then verified by oscillating disc rheometry and equilibrium swelling experiments. After critically examining various polymer types by X-ray photoelectron spectroscopy, electron spin resonance and Fourier transform infrared spectroscopy, a radical initiation mechanism was proposed to explain the self-crosslinking reaction. These NBR/graphite nanocomposites possess significantly improved wear resistance and gas barrier properties, and superior electrical/thermal conductivity. Such versatile functional properties make NBR nanocomposites a promising new class of advanced materials

  18. Improved mechanical and functional properties of elastomer/graphite nanocomposites prepared by latex compounding

    Energy Technology Data Exchange (ETDEWEB)

    Yang Jian [Key Laboratory for Nano-materials, Beijing University of Chemical Technology, Ministry of Education of China, Beijing 100029 (China); Key Laboratory on Preparation and Processing of Novel Polymer Materials, Beijing University of Chemical Technology, Beijing 100029 (China); Tian Ming [Key Laboratory for Nano-materials, Beijing University of Chemical Technology, Ministry of Education of China, Beijing 100029 (China); Jia Qingxiu [Key Laboratory on Preparation and Processing of Novel Polymer Materials, Beijing University of Chemical Technology, Beijing 100029 (China); Shi Junhong [Key Laboratory for Nano-materials, Beijing University of Chemical Technology, Ministry of Education of China, Beijing 100029 (China); Zhang Liqun [Key Laboratory for Nano-materials, Beijing University of Chemical Technology, Ministry of Education of China, Beijing 100029 (China); Key Laboratory on Preparation and Processing of Novel Polymer Materials, Beijing University of Chemical Technology, Beijing 100029 (China)], E-mail: zhanglq@mail.buct.edu.cn; Lim Szuhui; Yu Zhongzhen [Centre for Advanced Materials Technology (CAMT), School of Aerospace, Mechanical and Mechatronic Engineering (J07), University of Sydney, Sydney, NSW 2006 (Australia); Mai Yiuwing [Centre for Advanced Materials Technology (CAMT), School of Aerospace, Mechanical and Mechatronic Engineering (J07), University of Sydney, Sydney, NSW 2006 (Australia)], E-mail: y.mai@usyd.edu.au

    2007-10-15

    The facile latex approach has been adopted to finely incorporate graphite nanosheets into elastomeric polymer matrix to obtain high-performance elastomeric nanocomposites with improved mechanical properties and functional properties. Scanning electron microscopy, transmission electron microscopy and X-ray diffraction experiments show that the nanostructures of the final nanocomposites exhibit a high degree of exfoliation and intercalation of graphite in the nitrile-butadiene rubber (NBR) matrix. Mechanical and dynamic-mechanical tests demonstrate that the NBR/graphite nanocomposites possess greatly increased elastic modulus and tensile strength, and desirably strong interfaces. The unexpected self-crosslinking of elastomer/graphite nanocomposites was discovered and then verified by oscillating disc rheometry and equilibrium swelling experiments. After critically examining various polymer types by X-ray photoelectron spectroscopy, electron spin resonance and Fourier transform infrared spectroscopy, a radical initiation mechanism was proposed to explain the self-crosslinking reaction. These NBR/graphite nanocomposites possess significantly improved wear resistance and gas barrier properties, and superior electrical/thermal conductivity. Such versatile functional properties make NBR nanocomposites a promising new class of advanced materials.

  19. Mechanical and tribological properties of acrylonitrile–butadiene rubber filled with graphite and carbon black

    International Nuclear Information System (INIS)

    Wang, Lei Lei; Zhang, Li Qun; Tian, Ming

    2012-01-01

    Highlights: ► Graphite/carbon black/rubber micro- and nano-composites were prepared. ► Nanocomposites showed better mechanical properties and wear resistance. ► The effect of load and sliding speed on friction and wear is significant. ► Graphite lubricant film can reduce friction coefficient and wear rate. -- Abstract: In this work, acrylonitrile–butadiene rubber (NBR)/expanded graphite (EG)/carbon black (CB) micro- and nanocomposites were prepared by two different methods, and the resulting mechanical and tribological properties were compared with those of NBR/CB composites. Meanwhile, the effects of graphite dispersion and loading content, as well as the applied load and sliding velocity on the tribological behavior of the above composites under dry friction condition were also evaluated. The worn surfaces were analyzed by scanning electron microscopy (SEM) to disclose wear mechanism. As expected, the better the dispersion of graphite, the more remarkable enhancement on tensile and dynamic mechanical properties, and the greater reduction in the coefficient of friction (COF) and specific wear rate (W s ). It was found that a small amount of EG could effectively decrease COF and W s of NBR/CB composites because of the formation of graphite lubricant films. The COF and W s of NBR/CB/EG composites show a decreasing trend with a rise in applied load and sliding velocity. NBR/CB/EG nanocomposite always shows a stable wearing process with relatively low COF and W s . It is thought that well-dispersed graphite nano-sheets were beneficial to the formation of a fine and durable lubricant film.

  20. Development of electrically heated rods with resistive element of graphite or carbon/carbon composites for simulating transients in nuclear reactors

    International Nuclear Information System (INIS)

    Polidoro, H.A.

    1987-01-01

    Thermo-hydraulic problems, in nuclear plants are normally analysed by the use of electrically heated rods. The direct or indirect heater rods are limited in their use because, for high temperatures and high heat flux, the heating element temperature approach its melting point. The use of platinum or tantalum is not economically viable. Graphite and carbon/carbon composites are alternative materials because they are good electrical conductors and have good mechanical properties at high temperatures. Graphite and carbon/carbon composites were used to make heating elements for testing by indirect heating. The swaging process used to reduce the cladding diameter prevented the fabrication of graphite heater rods. Carbon/carbon composite used to make heating elements gave good results up to a heat flux of 100 W/cm 2 . It is easy to verify that this value can be exceeded if the choice of the complementary materials for insulator and cladding improved. (author) [pt

  1. Optical properties of NbCl5 and ZnMg intercalated graphite compounds

    International Nuclear Information System (INIS)

    Jung, Eilho; Lee, Seokbae; Roh, Seulki; Kang, Jihoon; Park, Tuson; Hwang, Jungseek; Meng, Xiuqing; Tongay, Sefaattin

    2014-01-01

    We studied NbCl 5 and ZnMg alloy intercalated graphite compounds using an optical spectroscopy technique. These intercalated metallic graphite samples were quite challenging to obtain optical reflectance spectra since they were not flat and quite thin. By using both a new method and an in situ gold evaporation technique we were able to obtain reliable reflectance spectra of our samples in the far and mid infrared range (80–7000 cm −1 ). We extracted the optical constants including the optical conductivity and the dielectric function from the measured reflectance spectra using a Kramers–Kronig analysis. We also extracted the dc conductivity and the plasma frequencies from the optical conductivity and dielectric functions. NbCl 5 intercalated graphite samples show similar optical conductivity spectra as bare highly oriented pyrolytic graphite even though there are some differences in detail. ZnMg intercalated samples show significantly different optical conductivity spectra from the bare graphite. Optical spectroscopy is one of the most reliable experimental techniques to obtain the electronic band structures of materials. The obtained optical conductivities support the recent theoretically calculated electronic band structures of NbCl 5 and ZnMg intercalated graphite compounds. Our results also provide important information of electronic structures and charge carrier properties of these two new intercalated materials for applications. (paper)

  2. Effect of Bi on graphite morphology and mechanical properties of heavy section ductile cast iron

    Directory of Open Access Journals (Sweden)

    Song Liang

    2014-03-01

    Full Text Available To improve the mechanical properties of heavy section ductile cast iron, bismuth (Bi was introduced into the iron. Five castings with different Bi content from 0 to 0.014 wt.% were prepared; and four positions in the casting from the edge to the center, with different solidification cooling rates, were chosen for microstructure observation and mechanical properties test. The effect of the Bi content on the graphite morphology and mechanical properties of heavy section ductile cast iron were investigated. Results show that the tensile strength, elongation and impact toughness at different positions in the five castings decrease with a decrease in cooling rate. With an increase in Bi content, the graphite morphology and the mechanical properties at the same position are improved, and the improvement of mechanical properties is obvious when the Bi content is no higher than 0.011wt.%. But when the Bi content is further increased to 0.014wt.%, the improvement of mechanical properties is not obvious due to the increase of chunky graphite number and the aggregation of chunky graphite. With an increase in Bi content, the tensile fracture mechanism is changed from brittle to mixture ductile-brittle fracture.

  3. Effect of microstructure on mechanical properties and machinability of spheroidal graphite cast iron

    International Nuclear Information System (INIS)

    Kubota, Satoru; Iio, Chinori; Yamaguchi, Shoji; Naito, Daiki; Tomota, Yo; Stefanus, Harjo

    2013-01-01

    Tensile properties, fatigue strength and machinability of spheroidal graphite cast irons with different microstructures were studied. Work-hardening and tensile strength increased with increasing pearlite volume fraction. In situ neutron diffraction during tensile deformation revealed that phase stresses and intergranular stresses are generated with deformation resulting in high work-hardening and high tensile strength with increasing pearlite volume fraction. It was found that graphite grains bear almost no stress, and strongly influence fatigue crack initiation as well as propagation. Therefore graphite refinement is very effective to realize high fatigue strength. The tool life for cutting becomes shorter with increasing pearlite volume fraction. The balance of mechanical properties and machinability was considered. (author)

  4. The tensile deformation behavior of nuclear-grade isotropic graphite posterior to hydrostatic loading

    International Nuclear Information System (INIS)

    Yoda, S.; Eto, M.

    1983-01-01

    The effects of prehydrostatic loading on microstructural changes and tensile deformation behavior of nuclear-grade isotropic graphite have been examined. Scanning electron micrographs show that formation of microcracks associated with delamination between basal planes occurs under hydrostatic loading. Hydrostatic loading on specimens results in the decrease in tensile strength and increase in residual strain generated by the applied tensile stress at various levels, indicating that the graphite material is weakened by hydrostatic loading. A relationship between residual strain and applied tensile stress for graphite hydrostatically-loaded at several pressure levels can be approximately expressed as element of= (AP + B) sigmasup(n) over a wide range hydrostatic pressure, where element of, P and sigma denote residual strain, hydrostatic pressure and applied tensile stress, respectively; A, B and n are constant. The effects of prehydrostatic loading on the tensile stress-strain behavior of the graphite were examined in more detail. The ratio of stress after hydrostatic loading to that before hydrostatic loading on the stress-strain relationship remains almost unchanged irrespective of strain. (orig.)

  5. Tritium retention properties of tungsten, graphite and co-deposited carbon film

    International Nuclear Information System (INIS)

    Nobuta, Y.; Hatano, Y.; Matsuyama, M.; Abe, S.; Akamaru, S.; Yamauchi, Y.; Hino, T.; Suzuki, S.; Akiba, M.

    2014-01-01

    DT + ion irradiation was performed on polycrystalline tungsten, graphite and carbon film and both the amount of retained tritium and the reduction of retained tritium after preservation in vacuum were investigated using an IP technique and BIXS. In addition, the relationship between the retention properties of tritium and the microstructure of graphite and carbon film were studied with Raman spectroscopy. The amount of retained tritium in tungsten was smaller than in both graphite and carbon film. After 1 keV of DT + irradiation, graphite showed no reduction of the amount of retained tritium after six months preservation while that of carbon film decreased by approximately 20% after 40 days preservation. It was suggested that this difference might be associated with differences in the microstructure between graphite and carbon film. In tungsten, the amount of retained tritium decreased to approximately half after 18 days preservation. As the incident energy of implanted tritium to tungsten increased, the decrease in tritium retention during preservation became slower. Tungsten's properties of releasing tritium while preserved in vacuum would be a useful tool for the reduction/removal of retained tritium

  6. 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.

  7. Growth mechanism and interface magnetic properties of Co nanostructures on graphite

    NARCIS (Netherlands)

    Wong, P.K.J.; de Jong, Machiel Pieter; Leonardus, L.; Siekman, Martin Herman; van der Wiel, Wilfred Gerard

    2011-01-01

    We investigated structural, electronic, and magnetic properties of Co adsorbed on highly oriented pyrolytic graphite (HOPG). Distribution and atomic sites of 3d transition-metal Co nanoislands and adatoms on HOPG were experimentally investigated by scanning tunneling microscopy with atomic

  8. Nuclear quadrupole interaction measurements of 19F* and 22Na* on Graphite

    International Nuclear Information System (INIS)

    Djoko-Surono, Th; Martin, Peter W

    1996-01-01

    Time differential perturbed angular distribution (TDPAD) technique has been used to investigate nuclear quadrupole interactions of 19 F * and 22 Na * in graphite. We concentrated the measurements on pseudo single crystal graphite called Highly Oriented Pyrolytic Graphite for it has an ordered structure in which the c-axes of the microcrystals aligned in a certain direction with the mosaic spread less than 1 o , while the a- and b-axes randomly oriented on a plane perpendicular to the c-axes. Interactions between quadrupole moment of 19 F * and 22 Na * with its surroundings electric field gradient were studied by detecting the γ-rays distribution, W(Θ,t). For 1 9F * we found one static interaction. The corresponding electric field gradient was V zz =3.24(19)x10 22 V/m 2 . In the case of 22 Na * we found no evidence of nuclear quadrupole interaction, however, we were able to conclude that |QV 22 | 19 bV/m 2 . Using theoretical calculation Q=0.06 barn, we find that |V zz | 20 V/m 2 . These results indicate that the value efg depend on two factors, the host crystal and the core electrons. The core electrons contribution to the total efg is considerably large

  9. 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.

  10. Electropolymerization of 3-aminophenol on carbon graphite surface: Electric and morphologic properties

    International Nuclear Information System (INIS)

    Franco, Diego L.; Afonso, Andre S.; Vieira, Sabrina N.; Ferreira, Lucas F.; Goncalves, Rafael A.; Brito-Madurro, Ana G.; Madurro, Joao M.

    2008-01-01

    This paper reports the formation of electropolymerized films derived from 3-aminophenol on graphite electrode by cyclic voltammetry, prepared in different pH conditions. With increase of pH values, a shift of the oxidation potential of 3-aminophenol to more cathodic potentials was observed. 3-Aminophenol electrooxidation, in acid and basic media, yielded polymeric films onto graphite surface. In ferrocyanide/ferricyanide solution, the polymer produced in acid medium showed higher electron transfer efficiency. Scanning electron microscopy (SEM), atomic force microscopy (AFM), and FT-IR were used to investigate some properties of the graphite electrode modified with poly(3-aminophenol). Scanning electron microscopy showed that the morphology of the films is strongly dependent on the pH of the electropolymerization medium. FT-IR spectra of polymer films produced for either acid or basic media suggest that the monomer is polymerized by NH 2 group

  11. Electropolymerization of 3-aminophenol on carbon graphite surface: Electric and morphologic properties

    Energy Technology Data Exchange (ETDEWEB)

    Franco, Diego L.; Afonso, Andre S.; Vieira, Sabrina N.; Ferreira, Lucas F. [Institute of Chemistry, Federal University of Uberlandia, Av. Joao Naves de Avila 2121, 38400-902 Uberlandia (Brazil); Goncalves, Rafael A. [School of Mechanical Engineering, Federal University of Uberlandia, Av. Joao Naves de Avila 2121, 38400-902 Uberlandia (Brazil); Brito-Madurro, Ana G. [Institute of Chemistry, Federal University of Uberlandia, Av. Joao Naves de Avila 2121, 38400-902 Uberlandia (Brazil); Madurro, Joao M. [Institute of Chemistry, Federal University of Uberlandia, Av. Joao Naves de Avila 2121, 38400-902 Uberlandia (Brazil)], E-mail: jmadurro@ufu.br

    2008-02-15

    This paper reports the formation of electropolymerized films derived from 3-aminophenol on graphite electrode by cyclic voltammetry, prepared in different pH conditions. With increase of pH values, a shift of the oxidation potential of 3-aminophenol to more cathodic potentials was observed. 3-Aminophenol electrooxidation, in acid and basic media, yielded polymeric films onto graphite surface. In ferrocyanide/ferricyanide solution, the polymer produced in acid medium showed higher electron transfer efficiency. Scanning electron microscopy (SEM), atomic force microscopy (AFM), and FT-IR were used to investigate some properties of the graphite electrode modified with poly(3-aminophenol). Scanning electron microscopy showed that the morphology of the films is strongly dependent on the pH of the electropolymerization medium. FT-IR spectra of polymer films produced for either acid or basic media suggest that the monomer is polymerized by NH{sub 2} group.

  12. Characterization of Epoxy Functionalized Graphite Nanoparticles and the Physical Properties of Epoxy Matrix Nanocomposites

    Science.gov (United States)

    Miller, Sandi G.; Bauer, Jonathan L.; Maryanski, Michael J.; Heimann, Paula J.; Barlow, Jeremy P.; Gosau, Jan-Michael; Allred, Ronald E.

    2010-01-01

    This work presents a novel approach to the functionalization of graphite nanoparticles. The technique provides a mechanism for covalent bonding between the filler and matrix, with minimal disruption to the sp2 hybridization of the pristine graphene sheet. Functionalization proceeded by covalently bonding an epoxy monomer to the surface of expanded graphite, via a coupling agent, such that the epoxy concentration was measured as approximately 4 wt.%. The impact of dispersing this material into an epoxy resin was evaluated with respect to the mechanical properties and electrical conductivity of the graphite-epoxy nanocomposite. At a loading as low as 0.5 wt.%, the electrical conductivity was increased by five orders of magnitude relative to the base resin. The material yield strength was increased by 30% and Young s modulus by 50%. These results were realized without compromise to the resin toughness.

  13. 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.

  14. 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

  15. Enhanced microwave absorption properties of graphite nanoflakes by coating hexagonal boron nitride nanocrystals

    KAUST Repository

    Zhong, Bo; Liu, Wei; Yu, Yuanlie; Xia, Long; Zhang, Jiulin; Chai, Zhenfei; Wen, Guangwu

    2017-01-01

    We report herein the synthesis of a novel hexagonal boron nitride nanocrystal/graphite nanoflake (h-BNNC/GNF) composite through a wet-chemistry coating of graphite nanoflakes and subsequent in-situ thermal treatment process. The characterization results of X-ray diffraction, scanning electron microscope, transmission electron microscope, energy dispersive X-ray spectrum, and X-ray photoelectron spectroscopy demonstrate that h-BNNCs with diameter of tens of nanometers are highly crystallized and anchored on the surfaces of graphite nanoflakes without obvious aggregation. The minimum reflection loss (RL) value of the h-BNNC/GNF based absorbers could reach −32.38dB (>99.99% attenuation) with the absorber thickness of 2.0mm. This result is superior to the other graphite based and some dielectric loss microwave absorption materials recently reported. Moreover, the frequency range where the RL is less than −10dB is 3.49-17.28GHz with the corresponding thickness of 5.0 to 1.5mm. This reveals a better electromagnetic microwave absorption performance of h-BNNC/GNFs from the X-band to the Ku-band. The remarkable enhancement of the electromagnetic microwave absorption properties of h-BNNC/GNFs can be assigned to the increase of multiple scattering, interface polarization as well as the improvement of the electromagnetic impedance matching of graphite nanoflakes after being coated with h-BNNCs.

  16. Enhanced microwave absorption properties of graphite nanoflakes by coating hexagonal boron nitride nanocrystals

    KAUST Repository

    Zhong, Bo

    2017-05-31

    We report herein the synthesis of a novel hexagonal boron nitride nanocrystal/graphite nanoflake (h-BNNC/GNF) composite through a wet-chemistry coating of graphite nanoflakes and subsequent in-situ thermal treatment process. The characterization results of X-ray diffraction, scanning electron microscope, transmission electron microscope, energy dispersive X-ray spectrum, and X-ray photoelectron spectroscopy demonstrate that h-BNNCs with diameter of tens of nanometers are highly crystallized and anchored on the surfaces of graphite nanoflakes without obvious aggregation. The minimum reflection loss (RL) value of the h-BNNC/GNF based absorbers could reach −32.38dB (>99.99% attenuation) with the absorber thickness of 2.0mm. This result is superior to the other graphite based and some dielectric loss microwave absorption materials recently reported. Moreover, the frequency range where the RL is less than −10dB is 3.49-17.28GHz with the corresponding thickness of 5.0 to 1.5mm. This reveals a better electromagnetic microwave absorption performance of h-BNNC/GNFs from the X-band to the Ku-band. The remarkable enhancement of the electromagnetic microwave absorption properties of h-BNNC/GNFs can be assigned to the increase of multiple scattering, interface polarization as well as the improvement of the electromagnetic impedance matching of graphite nanoflakes after being coated with h-BNNCs.

  17. Time of flight measurements of unirradiated and irradiated nuclear graphite under cyclic compressive load

    Energy Technology Data Exchange (ETDEWEB)

    Bodel, W., E-mail: william.bodel@hotmail.com [Nuclear Graphite Research Group, The University of Manchester (United Kingdom); Atkin, C. [Health and Safety Laboratory, Buxton (United Kingdom); Marsden, B.J. [Nuclear Graphite Research Group, The University of Manchester (United Kingdom)

    2017-04-15

    The time-of-flight technique has been used to investigate the stiffness of nuclear graphite with respect to the grade and grain direction. A loading rig was developed to collect time-of-flight measurements during cycled compressive loading up to 80% of the material's compressive strength and subsequent unloading of specimens along the axis of the applied stress. The transmission velocity (related to Young's modulus), decreased with increasing applied stress; and depending on the graphite grade and orientation, the modulus then increased, decreased or remained constant upon unloading. These tests were repeated while observing the microstructure during the load/unload cycles. Initial decreases in transmission velocity with compressive load are attributed to microcrack formation within filler and binder phases. Three distinct types of behaviour occur on unloading, depending on the grade, irradiation, and loading direction. These different behaviours can be explained in terms of the material microstructure observed from the microscopy performed during loading.

  18. Research activities of the nuclear graphite research group at the University of Manchester, UK

    International Nuclear Information System (INIS)

    Marsden, B.J.; Fok, A.S.L.; Marrow, J.; Mummery, P.

    2004-01-01

    In 2001 the Nuclear Safety Division (NSD) of the UK Health and Safety Executive (HSE) decided to underwrite the Nuclear Graphite Research Group (NGRG) at the University of Manchester, UK with the aim of providing a source of independent research and advice to the HSE (NSD). Since then the group has rapidly expanded to 16 members and attracted considerable funding from the nuclear power industry and the regulator for a wide range of research and consultancy work. It is now also part of the Material Performance Centre within the BNFL Universities Research Alliance. Extensive collaboration exists between the group and other nuclear research institutes, both in the UK and overseas. This paper briefly describes some of the research programmes being carried out by the NGRG at Manchester. (author)

  19. 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

  20. 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.

  1. Experimental investigation of thermophysical properties of eutectic Mo–C, graphite and tantalum at high temperatures

    International Nuclear Information System (INIS)

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

    2016-01-01

    An experimental technique based on fast electrical heating for investigation of thermophysical properties of refractory materials under high pressures and at high temperatures is considered. A set of thermophysical properties of refractory materials such as specific enthalpy, specific heat capacity, specific resistivity, melting heat of eutectic Mo-C and thermal expansion of graphite and tantalum were determined. The obtained temperature of eutectic melting of MoC 0.82 shows close agreement with equilibrium Mo-C phase diagram. (paper)

  2. Effect of graphite loading on the electrical and mechanical properties of Poly (Ethylene Oxide)/Poly (Vinyl Chloride) polymer films

    Science.gov (United States)

    Hajar, M. D. S.; Supri, A. G.; Hanif, M. P. M.; Yazid, M. I. M.

    2017-10-01

    In this study, films consisting of a blend of poly (ethylene oxide)/poly (vinyl chloride) (PEO/PVC) and a conductive filler, graphite were prepared and characterized for their mechanical and electrical properties. Solid polymer blend films based on PEO/PVC (50/50 wt%/wt%) with different graphite loading were prepared by using solution casting technique. Electrical conductivity results discovered the conductivity increased with increasing of filler loading. However, increasing amount of graphite loading led to a decreased in tensile strength and young’s modulus of PEO/PVC/Graphite polymer films. The dispersion of graphite and mechanism of conductive path in the polymer films were also investigated by scanning electron microscopy (SEM). The morphology of the PEO/PVC/Graphite polymer films shows that agglomeration occurred to complete the connection of conductive path, thus improving the conductivity behavior of the polymer films.

  3. Change in physical properties of high density isotropic graphites irradiated in the ?JOYO? fast reactor

    Science.gov (United States)

    Maruyama, T.; Kaito, T.; Onose, S.; Shibahara, I.

    1995-08-01

    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 2.86 × 10 26 n/m 2 ( E > 0.1 MeV) at temperatures from 549 to 597°C. Postirradiation examination was carried out on the dimensional changes, elastic modulus, and thermal conductivity of these materials. Dimensional change results indicate 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 by two to three times the unirradiated values. The large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependence on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, with a negligible change in specific heat. 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.

  4. Change in physical properties of high density isotropic graphites irradiated in the ''JOYO'' fast reactor

    International Nuclear Information System (INIS)

    Maruyama, T.; Kaito, T.; Onose, S.; Shibahara, I.

    1995-01-01

    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 597 C. Postirradiation examination was carried out on the dimensional changes, elastic modulus, and thermal conductivity of these materials. Dimensional change results indicate 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 by two to three times the unirradiated values. The large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependence on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, with a negligible change in specific heat. 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. (orig.)

  5. Effect of mechanical treatment on the electrical properties of graphite materials

    International Nuclear Information System (INIS)

    Yunasfi; Salim Mustofa

    2010-01-01

    Measurement of electrical properties of graphite materials as the mechanical treatment result with high energy milling (HEM) techniques was carried out. The carbon powder was milled using HEM by varying the milling time from 25 hours to 100 hours, and afterwards the graphite material was formed to a pellet under compaction up to 20 ton/cm"2. The measurements result of electrical properties using LCR (Inductance, Capacitance, Resistance) meter showed that milling process to graphite causes the increasing of conductivity and capacitance values and these values were increasing with the increase of milling times. Before milling, the conductivity value is 3.5976 Siemens/cm and the value of capacitance is 0.2223 μF at 100 kHz frequency. Increasing of conductivity value reached 26 % and capacitance value reached 66 % after milling for 100 hours. These increasing are analyzed due to decreasing of graphite powder particle size causing by longer milling time and result in higher electrical conductivity. (author)

  6. Electrical properties of alkali-activated slag composite with combined graphite/CNT filler

    Science.gov (United States)

    Rovnaník, P.; Míková, M.; Kusák, I.

    2017-10-01

    Alkali-activated industrial by-products such as blast furnace slag are known to possess properties which are comparable to or even better than those observed for ordinary Portland cement. The combination of alkali-activated slag matrix with conductive filler introduces new functionalities which are commonly known for self-sensing or self-heating concrete. The present paper discusses the effect of the mixture of two different conductive fillers, graphite powder and carbon nanotubes (CNTs), on the electrical properties of alkali-activated slag mortars. Prepared samples were also tested for their mechanical properties and microstructure was investigated by means of mercury intrusion porosimetry and scanning electron microscopy. The percolation threshold for the resistance was reached for the mixture containing 0.1% CNTs and 8% graphite powder.

  7. 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.)

  8. Transport Properties in Nuclear Pasta

    Science.gov (United States)

    Caplan, Matthew; Horowitz, Charles; Berry, Donald; da Silva Schneider, Andre

    2016-09-01

    At the base of the inner crust of neutron stars, where matter is near the nuclear saturation density, nuclear matter arranges itself into exotic shapes such as cylinders and slabs, called `nuclear pasta.' Lepton scattering from these structures may govern the transport properties of the inner crust; electron scattering from protons in the pasta determines the thermal and electrical conductivity, as well as the shear viscosity of the inner crust. These properties may vary in pasta structures which form at various densities, temperatures, and proton fractions. In this talk, we report on our calculations of lepton transport in nuclear pasta and the implication for neutron star observables.

  9. Controlled synthesis of graphitic carbon-encapsulated α-Fe2O3 nanocomposite via low-temperature catalytic graphitization of biomass and its lithium storage property

    International Nuclear Information System (INIS)

    Wu, Feng; Huang, Rong; Mu, Daobin; Wu, Borong; Chen, Yongjian

    2016-01-01

    Highlights: • Facile synthesis of graphitic carbon/α-Fe 2 O 3 nano-sized anode composite. • In situ low temperature catalytic graphitization of biomass material. • Onion-like graphitic carbon layers conformally encapsulating around α-Fe 2 O 3 core. • High lithium storage properties, especially, outstanding cycle performance. - Abstract: A delicate structure of graphitic carbon-encapsulated α-Fe 2 O 3 nanocomposite is in situ constructed via “Absorption–Catalytic graphitization–Oxidation” strategy, taking use of biomass matter of degreasing cotton as carbon precursor and solution reservoir. With the assistance of the catalytic graphitization effect of iron core, onion-like graphitic carbon (GC) shell is made directly from the biomass at low temperature (650 °C). The nanosized α-Fe 2 O 3 particles would effectively mitigate volumetric strain and shorten Li + transport path during charge/discharge process. The graphitic carbon shells may promote charge transfer and protect active particles from directly exposing to electrolyte to maintain interfacial stability. As a result, the as-prepared α-Fe 2 O 3 @GC composite displays an outstanding cycle performance with a reversible capacity of 1070 mA h g −1 after 430 cycles at 0.2C, as well as a good rate capability of ∼ 950 mA h g −1 after 100 cycles at 1C and ∼ 850 mA h g −1 even up to 200 cycles at a 2C rate.

  10. AGC-3 Graphite Preirradiation Data Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    William Windes; David Swank; David Rohrbaugh; Joseph Lord

    2013-09-01

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

  11. 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

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

    International Nuclear Information System (INIS)

    Fitzer, E.; Koechling, K.H.

    1984-01-01

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

  13. 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

  14. Physical Properties of 3D Interconnected Graphite Networks - Aerographite

    Science.gov (United States)

    2015-10-30

    resistance could be observed. The dielectric properties are studied using AC impedance spectroscopy. A low capacitance can be measured that refers to...current frequency ..................................................... 17 Figure 3.11: Parallel circuit of resistance and capacitor (left) and...release; distribution is unlimited. Figure 3.11: Parallel circuit of resistance and capacitor (left) and Nyquist plot of reactance against re- sistance

  15. Synthesis and photocatalytic properties of graphitic carbon nitride nanofibers using porous anodic alumina templates

    Science.gov (United States)

    Suchitra, S. M.; Udayashankar, N. K.

    2017-12-01

    In the present study, we describe an effective method for the synthesis of Graphitic carbon nitride (GCN) nanostructures using porous anodic alumina (AAO) membrane as template by simple thermal condensation of cyanamide. Synthesized nanostructure was fully analysed by various techniques to detect its crystalline nature, morphology, luminescent properties followed by the evaluation of its photocatalytic activity in the degradation of Methylene blue dye. Structural analysis of synthesized GCNNF was systematically carried out using x-ray powder diffraction (XRD) and scanning electron microscope (SEM), and. The results confirmed the growth of GCN inside the nanochannels of anodic alumina templates. Luminescent properties of GCNNF were studied using photoluminescence (PL) spectroscopy. PL analysis showed the presence of a strong emission peak in the wavelength range of 350-600 nm in blue region. GCNNF displays higher photocatalytic performance in the photodegradation of methylene blue compare to the bulk GCN. Highlights 1. In the present paper, we report the synthesis of graphitic carbon nitride nanofibers (GCNNF) using porous anodic aluminium oxide membranes as templates through thermal condensation of cyanamide at 500 °C. 2. The synthesis of Graphitic carbon nitride nanofibers using porous andic alumina template is the efficient approach for increasing crystallinity and surface area. 3. The high surface area of graphitic carbon nitride nanofibers has a good impact on novel optical and photocatalytic properties of the bulkGCN. 4. AAO templating of GCN is one of the versatile method to produce tailorable GCN nanostructures with higher surface area and less number of structural defects. 5. Towards photocatalytic degradation of dyes, the tuning of physical properties is very essential thing hence we are succeeded in achieving better catalytic performance of GCN nanostructures by making use of AAO templates.

  16. Artificial graphites

    International Nuclear Information System (INIS)

    Maire, J.

    1984-01-01

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

  17. 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.

  18. 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.

  19. Influence of expanded graphite (EG and graphene oxide (GO on physical properties of PET based nanocomposites

    Directory of Open Access Journals (Sweden)

    Paszkiewicz Sandra

    2014-12-01

    Full Text Available This work is the continuation and refinement of already published communications based on PET/EG nanocomposites prepared by in situ polymerization1, 2. In this study, nanocomposites based on poly(ethylene terephthalate with expanded graphite were compared to those with functionalized graphite sheets (GO. The results suggest that the degree of dispersion of nanoparticles in the PET matrix has important effect on the structure and physical properties of the nanocomposites. The existence of graphene sheets nanoparticles enhances the crystallization rate of PET. It has been confirmed that in situ polymerization is the effective method for preparation nanocomposites which can avoid the agglomeration of nanoparticles in polymer matrices and improve the interfacial interaction between nanofiller and polymer matrix. The obtained results have shown also that due to the presence of functional groups on GO surface the interactions with PET matrix can be stronger than in the case of exfoliated graphene (EG and matrix.

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

    Science.gov (United States)

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

    2009-09-01

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

  1. Mechanical properties of low-density polyethylene filled by graphite nanoplatelets

    DEFF Research Database (Denmark)

    Carotenuto, G.; De Nicola, S.; Palomba, M.

    2012-01-01

    The mechanical properties of GNP/LDPE nanocomposites (graphite nanoplatelets/low density polyethylene) have been investigated, in order to establish the effect of nanoscale reinforcement within the polymer matrix. Results show that the presence of the filler does not involve a change...... in the microscopic structure of the polymer. However, on a macroscopic scale, GNPs limit the mobility of the polymer chains, resulting in an increase in stiffness for the final composite. Orientation of GNPs within the LDPE matrix is also an important issue that affects mechanical properties and it has been...

  2. New graphite/salt materials for high temperature energy storage. Phase change properties study

    International Nuclear Information System (INIS)

    Lopez, J.

    2007-07-01

    This work is a contribution to the study of new graphite/salt composites dedicated to high temperature energy storage (≥200 C). The aim is to analyse and to understand the influence of both graphite and composite microstructure on the phase change properties of salts. This PhD is carried out within the framework of two projects: DISTOR (European) and HTPSTOCK (French). The major contributions of this work are threefold: 1) An important database (solid-liquid phase change properties) is provided from the DSC analysis of six salts and the corresponding composites. 2) Rigorous modeling of salts melting in confined media in several geometries are proposed to understand why, during the first melting of the compression elaborated composites, problems of salt leakage are observed. These models show that the materials morphology is responsible for these phenomena: the graphite matrix restrains the volume expansion due to salt melting: salt melts under pressure, which leads to a melting on a large temperature range and to a loss of energy density. Sensitivity analysis of parameters (geometric and physic) shows that matrix rigidity modulus is the parameter on which it is necessary to act during the composites elaboration to blur this phenomenon. 3) Finally, this work proposes a thermodynamic formulation of both surface/interface phenomena and the presence of dissolved impurities being able to explain a melting point lowering. It seems that the melting point lowering observed (∼ 5 C) are mainly due to the presence of dissolved impurities (brought by graphite) in the liquid, along with an additional Gibbs-Thomson effect (∼ 1 C, related to the size of the clusters crystals). (author)

  3. 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)

  4. Characterization of {sup 14}C in neutron irradiated NBG-25 nuclear graphite

    Energy Technology Data Exchange (ETDEWEB)

    LaBrier, Daniel, E-mail: labrdani@isu.edu; Dunzik-Gougar, Mary Lou

    2014-05-01

    Recent studies suggest that the highest concentration of {sup 14}C contamination present in reactor-irradiated graphite exists on the surfaces and within near-surface layers. Surface-sensitive analysis techniques (XPS, ToF-SIMS, SEM/EDS and Raman) were employed to determine the chemical nature of {sup 14}C on irradiated NBG-25 (nuclear grade) graphite surfaces. Several {sup 14}C precursor species are identified on the surfaces of irradiated NBG-25; the quantities of these species decrease at sub-surface depths, which further suggests that {sup 14}C formation is predominantly a surface-concentrated phenomenon. The elevated presence of several surface oxide complexes on irradiated NBG-25 surfaces are attributed directly to neutron irradiation. Larger numbers of oxide bonds were found on irradiated NBG-25 surfaces (when compared to unirradiated samples) in the form of interlattice (e.g. ether) and dangling (e.g. carboxylate and ketone) bonds; the quantities of these bond types also decrease with increasing sub-surface depths.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

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

  6. Testing of different data libraries in activation analysis of concrete and graphite from nuclear installations

    International Nuclear Information System (INIS)

    Cometto, M.; Ancius, D.; Ridikas, D.

    2003-01-01

    With the aging of the nuclear park, decommissioning and dismantling of nuclear installations after their service life is becoming an important issue for the nuclear industry. The radiological characterisation of the equipment and structures present in the reactor and its environment is an essential stage in a decommissioning project since it permits to define and optimize the decommissioning strategy and the disassembling operations. In addition, correct activation estimates are essential for determining the quantity and the nature of the radiological waste generated during decommissioning. The adoption of efficient dismantling procedures and the optimization of the mass flow going to different waste repositories might reduce substantially the total cost of decommissioning. The present work has been done in the framework of the decommissioning and dismantling of the experimental reactor of the University of Strasbourg (RUS). A methodology that combines theoretical calculations and direct measurements has been developed for determining the long-term induced activity in the graphite, concrete and materials present in the reactor. After characterisation of the different elements present in the reactor, it is then possible to plan efficiently the disassembling and dismantling of the system and to optimise the mass flow going to different waste repositories. From a scientific perspective, the comparison of theoretical predictions with experimental values validates the approach and the methodology used in the present study and tests the consistency and the reliability of the nuclear data used for activation analysis. (orig.)

  7. Graphite to Graphene via Graphene Oxide: An Overview on Synthesis, Properties, and Applications

    Science.gov (United States)

    Hansora, D. P.; Shimpi, N. G.; Mishra, S.

    2015-12-01

    This work represents a state-of-the-art technique developed for the preparation of graphene from graphite-metal electrodes by the arc-discharge method carried out in a continuous flow of water. Because of continuous arcing of graphite-metal electrodes, the graphene sheets were observed in water with uniformity and little damage. These nanosheets were subjected to various purification steps such as acid treatment, oxidation, water washing, centrifugation, and drying. The pure graphene sheets were analyzed using Raman spectrophotometry, x-ray diffraction (XRD), field emission-scanning electron microscopy (FE-SEM), and tunneling electron microscopy (TEM). Peaks of Raman spectra were recorded at (1300-1400 cm-1) and (1500-1600 cm-1) for weak D-band and strong G-band, respectively. The XRD pattern showed 85.6% crystallinity of pure graphite, whereas pure graphene was 66.4% crystalline. TEM and FE-SEM micrographs revealed that graphene sheets were overlapped to each other and layer-by-layer formation was also observed. Beside this research work, we also reviewed recent developments of graphene and related nanomaterials along with their preparations, properties, functionalizations, and potential applications.

  8. Adhesive properties and adhesive joints strength of graphite/epoxy composites

    Science.gov (United States)

    Rudawska, Anna; Stančeková, Dana; Cubonova, Nadezda; Vitenko, Tetiana; Müller, Miroslav; Valášek, Petr

    2017-05-01

    The article presents the results of experimental research of the adhesive joints strength of graphite/epoxy composites and the results of the surface free energy of the composite surfaces. Two types of graphite/epoxy composites with different thickness were tested which are used to aircraft structure. The single-lap adhesive joints of epoxy composites were considered. Adhesive properties were described by surface free energy. Owens-Wendt method was used to determine surface free energy. The epoxy two-component adhesive was used to preparing the adhesive joints. Zwick/Roell 100 strength device were used to determination the shear strength of adhesive joints of epoxy composites. The strength test results showed that the highest value was obtained for adhesive joints of graphite-epoxy composite of smaller material thickness (0.48 mm). Statistical analysis of the results obtained, the study showed statistically significant differences between the values of the strength of the confidence level of 0.95. The statistical analysis of the results also showed that there are no statistical significant differences in average values of surface free energy (0.95 confidence level). It was noted that in each of the results the dispersion component of surface free energy was much greater than polar component of surface free energy.

  9. Specific features of the electrical properties in partially graphitized porous biocarbons of beech wood

    Science.gov (United States)

    Popov, V. V.; Orlova, T. S.; Gutierrez-Pardo, A.; Ramirez-Rico, J.

    2015-09-01

    The electrical and galvanomagnetic properties of partially graphitized highly porous bioC(Ni) biocarbon matrices produced by pyrolysis (carbonization) of beech wood at temperatures T carb = 850-1600°C in the presence of a Ni-containing catalyst have been studied in comparison with their microstructural features. The temperature dependences of the resistivity, the magnetoresistance, and the Hall coefficient have been measured in the temperature range of 4.2-300 K in magnetic fields to 28 kOe. It has been shown that an additional graphite phase introduction into samples with T carb ≥ 1000°C results in an increase in the carrier mobility by a factor of 2-3, whereas the carrier (hole) concentration remains within ~1020 cm-3, as in biocarbons obtained without catalyst. An analysis of experimental data has demonstrated that the features of the conductivity and magnetoresistance of these samples are described by quantum corrections related to their structural features, i.e., the formation of a globular graphite phase of nano- and submicrometer sizes in the amorphous matrix. The quantum corrections to the conductivity decrease with increasing carbonization temperature, which indicates an increase in the degree of structure ordering and is in good agreement with microstructural data.

  10. Electronic and spectroscopic properties of early 3d metal atoms on a graphite surface

    Science.gov (United States)

    Rakotomahevitra, A.; Garreau, G.; Demangeat, C.; Parlebas, J. C.

    1995-07-01

    High-sensitivity magneto-optic Kerr effect experiments failed to detect manifestations of magnetism in epitaxial films of V on Ag(100) substrates. More recently V 3s XPS of freshly evaporated V clusters on graphite exhibited the appearance of a satellite structure which has then been interpreted by the effect of surface magnetic moments on V. It is the absence of unambiguous results on the electronic properties of early 3d supported metals that prompts us to examine the problem. Our purpose is twofold. In a first part, after a total energy calculation within a tight-binding method which yields the equilibrium position of a given adatom, we use the Hartree-Fock approximation to find out a possible magnetic solution of V (or Cr) upon graphite for a reasonable value of the exchange integral Jdd. In a second part the informations given by the density of states of the graphite surface as well as the additional states of the adsorbed atom are taken into account through a generalised impurity Anderson Hamiltonian which incorporates the various Coulomb and exchange interactions necessary to analyse the 3s XPS results.

  11. 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.

  12. Mechanical property characterization and impact resistance of selected graphite/PEEK composite materials

    Science.gov (United States)

    Baker, Donald J.

    1994-01-01

    To use graphite polyetheretherketone (PEEK) material on highly curved surfaces requires that the material be drapable and easily conformable to the surface. This paper presents the mechanical property characterization and impact resistance results for laminates made from two types of graphite/PEEK materials that will conform to a curved surface. These laminates were made from two different material forms. These forms are: (1) a fabric where each yarn is a co-mingled Celion G30-500 3K graphite fiber and PEEK thermoplastic fiber; and (2) an interleaved material of Celion G30-500 3K graphite fabric interleaved with PEEK thermoplastic film. The experimental results from the fabric laminates are compared with results for laminates made from AS4/PEEK unidirectional tape. The results indicate that the tension and compression moduli for quasi-isotropic and orthotropic laminates made from fabric materials are at least 79 percent of the modulus of equivalent laminates made from tape material. The strength of fabric material laminates is at least 80 percent of laminates made from tape material. The evaluation of fabric material for shear stiffness indicates that a tape material laminate could be replaced by a fabric material laminate and still maintain 89 percent of the shear stiffness of the tape material laminate. The notched quasi-isotropic compression panel failure strength is 42 to 46 percent of the unnotched quasi-isotropic laminate strength. Damage area after impact with 20 ft-lbs of impact energy is larger for the co-mingled panels than for the interleaved panels. The inerleaved panels have less damage than panels made from tape material. Residual compression strength of quasi-isotropic panels after impact of 20 ft-lbs of energy varies between 33 percent of the undamaged quasi-isotropic material strength for the tape material and 38 percent of the undamaged quasi-isotropic material strength for the co-mingled fabric material.

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

    International Nuclear Information System (INIS)

    Morgan, W.C.

    1996-01-01

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

  14. Properties of the nuclear medium

    International Nuclear Information System (INIS)

    Baldo, M; Burgio, G F

    2012-01-01

    We review our knowledge on the properties of the nuclear medium that have been studied, over many years, on the basis of many-body theory, laboratory experiments and astrophysical observations. Throughout the presentation particular emphasis is placed on the possible relationship and links between the nuclear medium and the structure of nuclei, including the limitations of such an approach. First we consider the realm of phenomenological laboratory data and astrophysical observations and the hints they can give on the characteristics that the nuclear medium should possess. The analysis is based on phenomenological models, that however have a strong basis on physical intuition and an impressive success. More microscopic models are also considered, and it is shown that they are able to give invaluable information on the nuclear medium, in particular on its equation of state. The interplay between laboratory experiments and astrophysical observations is particularly stressed, and it is shown how their complementarity enormously enriches our insights into the structure of the nuclear medium. We then introduce the nucleon–nucleon interaction and the microscopic many-body theory of nuclear matter, with a critical discussion about the different approaches and their results. The Landau–Fermi liquid theory is introduced and briefly discussed, and it is shown how fruitful it can be in discussing the macroscopic and low-energy properties of the nuclear medium. As an illustrative example, we discuss neutron matter at very low density, and it is shown how it can be treated within the many-body theory. The general bulk properties of the nuclear medium are reviewed to indicate at which stage of our knowledge we stand, taking into account the most recent developments both in theory and experiments. A section is dedicated to the pairing problem. The connection with nuclear structure is then discussed, on the basis of the energy density functional method. The possibility of

  15. THE INFLUENCE OF RECUPIRATION’S METHODS OF GRAPHITE TO PROPERTIES OF SYNTHESIZED DIAMONDS

    Directory of Open Access Journals (Sweden)

    G. P. Bogatyreva

    2015-03-01

    Full Text Available The graphite’s waste can be used for synthesis of diamonds. It is established, that activation of graphite’s waste essential influence on a degree of transition of graph-ite in diamond and their physico-chemical properties. The activation of th graphite’s waste changes essentially their absorption and structural characteristics and to a great extent affect the characteristics of synthesized diamond. Thermal activation of graphite’s waste leads to that are synthesized, basically, diamond micropowders, and electrochemical — diamond grinding powders.

  16. Nuclear properties from perturbative QCD

    International Nuclear Information System (INIS)

    Close, F.E.; Roberts, R.G.; Ross, G.G.

    1986-01-01

    Two apparently different descriptions of quark distributions in a nucleus may in fact be connected. A ''duality'' between the QCD approach and the conventional model of nucleon binding leads to nuclear properties being simply related to the anomalous dimensions of QCD. (orig.)

  17. Development of mercury porosimeter. Application to nuclear graphite studies (1961); Mise au point d'un porosimetre a mercure. Application a l'etude des graphites nucleaires (1961)

    Energy Technology Data Exchange (ETDEWEB)

    Bocquet, M; Genisson, J; Sailleau, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1961-07-01

    A mercury porosimeter, model IFP-CEA, has been developed for application to nuclear graphite studies. The apparatus is based on the capillary depression phenomenon. The relationship between the radius of a pore and the pressure at which mercury fills it is pr = -2 {sigma} cos {theta} ( {sigma} is the surface tension, {theta} the angle of contact of the mercury). After some theoretical consideration, the apparatus is described. The mercury pressure is increased step-wise from 0 to 1000 kg/cm{sup 2} thus yielding the complete distribution of pores from 92 {mu} to 75 A. Results are, then presented concerning nuclear graphites which show the evolution of the porous structure under the effect of bitumen impregnation. In general, the volume of the large pores decreases while that of the small pores increases. The structure of impregnated products appears to depend to a certain extent on that of the starting materials. It has also been possible to study other products with this porosimeter; the range of measurements possible is such that it may be used for the study of the majority of porous materials. (authors) [French] Un porosimetre a mercure modele IFP-CEA a ete mis au point en vue de son application a l'etude de graphites nucleaires. Le fonctionnement de l'appareil repose sur le phenomene de depression capillaire. On etablit la relation existant entre le rayon r d'un pore et la pression p pour laquelle le mercure peut y penetrer: pr = -2 {sigma} cos {theta} ( {sigma} tension superficielle, {theta} angle de contact du mercure). Apres quelques considerations theoriques, l'appareil utilise est decrit. Il permet de faire varier par palier la pression du mercure entre 0 et 1000 kg/cm{sup 2} et d'etablir ainsi la distribution complete des rayons de pores compris entre 92 {mu} et 75 A. Les resultats d'une etude faite sur des graphites nucleaires sont alors presentes faisant apparaitre l'evolution de la structure poreuse sous l'effet des impregnations au brai. D'une facon

  18. The characteristics of TiC and oxidation resistance and mechanical properties of TiC coated graphite under corrosive environment

    International Nuclear Information System (INIS)

    Yoda, Shinichi; Oku, Tatsuo; Ioka, Ikuo; Umekawa, Shokichi.

    1982-07-01

    Core region of the Very High Temperature Gas Cooled Reactor (VHTR) consists mainly of polycrystalline graphite whose mechanical properties degradated by corrosion resulting from such impurities as O 2 , H 2 O, and CO 2 in coolant He gas. Mechanical properties and oxidation resistance of TiC coated graphite under corrosive condition were examined in order to evaluate the effects of TiC coating on preventing the graphite from its degradation in service condition of the VHTR. Characteristics of TiC coating was also examined using EPMA. Holding the specimen at 1373 K for 6 hr produced strong interface between TiC coating and the graphite, however, microcracks on TiC coating was observed, the origin of which is ascribed to mismatch in thermal expansion between TiC coating and the graphite. Oxidation rate of TiC coated graphite was one-thirds of that of uncoated graphite, which demonstrated that TiC coating on the graphite improved the oxidation resistance of the graphite. However, debonding of TiC coating layer at the interface was observed after heating for 3 to 4 hr in the oxidation condition. Changes in Young's modulus of TiC coated graphite were a half of that of uncoated graphite. Flexural strength of TiC coated graphite remained at the original value up to about 4 hr oxidation, therafter it decreased abruptly as was the trend of uncoated graphite. It is concluded that TiC coating on graphite materials is very effective in improving oxidation resistance and suppressing degradation of mechanical properties of the graphite. (author)

  19. 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

    difference in the two streams lays in the chosen retrieval technology, reflecting the different problems involved. The most significant differences in the retrieval solutions are; - MAC recovery is achieved with a hydraulic Artisan manipulator arm deployed through a series of new access points in the vault roof; - FED recovery utilizes two retrieval units which deploy petal grabs through tubes originally connected to the waste discharge conveyors. The waste is then transported for further treatment including monitoring, packing, lid and grouting fitting, box filling and swabbing of the boxes to eliminate external contamination. This poster session discusses the experiences made with the retrieval and emptying of graphite containing nuclear waste silos

  20. Standard Test Methods for Properties of Continuous Filament Carbon and Graphite Fiber Tows

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1999-01-01

    1.1 These test methods cover the preparation and tensile testing of resin-impregnated and consolidated test specimens made from continuous filament carbon and graphite yarns, rovings, and tows to determine their tensile properties. 1.2 These test methods also cover the determination of the density and mass per unit length of the yarn, roving, or tow to provide supplementary data for tensile property calculation. 1.3 These test methods include a procedure for sizing removal to provide the preferred desized fiber samples for density measurement. This procedure may also be used to determine the weight percent sizing. 1.4 These test methods include a procedure for determining the weight percent moisture adsorption of carbon or graphite fiber. 1.5 The values stated in SI units are to be regarded as the standard. The values in parentheses are for information only. 1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of t...

  1. Comparison of 3 MeV C+ Ion-Irradiation Effects between The Nuclear Graphites made of Pitch and Petroleum Cokes

    International Nuclear Information System (INIS)

    Se-Hwan, Chi; Gen-Chan, Kim; Jong-Hwa, Chang

    2006-01-01

    Currently, all the commercially available nuclear graphite grades are being made from two different cokes, i.e., petroleum coke or coal-tar pitch coke, and a coal-tar pitch binder. Of these, since the coke composes most of the graphite volume, i.e., > 70 %, it is understood that a physical, chemical, thermal, and mechanical property as well as an irradiation-induced property change will be strongly dependent on the type of coke. To obtain first-hand information on the effects of the coke type, i.e., petroleum or pitch, on the irradiation sensitivity of graphite, specimens made of IG-110 of petroleum coke and IG-430 of pitch coke were irradiated up to ∼ 19 dpa by 3 MeV C + at room temperature, and the irradiation-induced changes in the hardness, Young's modulus, Raman spectrum, and oxidation properties were characterized. Results of the TEM show that the size and density of the Mrozowski cracks appeared to be far larger and higher in the IG-110 than the IG-430. Results of the hardness test revealed a slightly higher increase in the IG-430 than the IG-110 by around 10 dpa, and the Raman spectrum measurement showed a higher (FWHM) D /(FWHM) G value for IG-430 for 0.02 ∼ 0.25 dpa. Both the hardness and Raman measurement may imply a higher irradiation sensitivity of the IG-430 than the IG-110. Results of the Young's modulus measurements showed a large data scattering, which prevented us from estimating the differences between the grades. Oxidation experiments using a TG-DTA under a flow of dry air/He = 2.5 % (flow rate: 40 CC/min) at 750 and 1000 deg C show that the IG-110 of the petroleum coke exhibits a far higher oxidation rate than the IG-430. The discrepancy between the oxidation rate of the two grades increased with an increase in the oxidation temperature and the dose. Oxidized surface pore area was larger for IG-110. Judging from the results obtained from the present experimental conditions, the irradiation sensitivity appeared to be dependent on the degree

  2. 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

  3. Mechanical properties of aluminium based metal matrix composites reinforced with graphite nanoplatelets

    Energy Technology Data Exchange (ETDEWEB)

    Alam, Syed Nasimul, E-mail: syedn@nitrkl.ac.in; Kumar, Lailesh

    2016-06-14

    In this work Al-matrix composites reinforced by exfoliated graphite nanoplatelets (xGnP) is fabricated by powder metallurgy route and their microstructure, mechanical properties and sliding wear behaviour were investigated. Here, xGnP has been synthesized from the thermally exfoliated graphite produced from a graphite intercalation compound (GIC) through rapid evaporation of the intercalant at an elevated temperature. The xGnP synthesized was characterized using scanning electron microscope (SEM), high-resolution transmission electron microscope (HRTEM), x-ray diffraction (XRD), atomic force microscopy (AFM), x-ray photoelectron spectroscopy (XPS), differential scanning calorimetry and thermogravimetric analysis (DSC/TGA), Raman spectroscopy and Fourier transform infrared spectroscopy (FTIR). The Al and xGnP powder mixtures were consolidated under a load of 565 MPa followed by sintering at 550 °C for 2 h in an inert atmosphere. Al-1, 2, 3 and 5 wt% xGnP nanocomposites were developed. Results of the wear test show that there was a significant improvement in the wear resistance of the composites up to the addition of 3 wt% of xGnP in the Al matrix. The hardness of the various Al-xGnP composites also shows improvement upto the addition of 1 wt% xGnP beyond which there was a decrease in the hardness of the composites. The tensile strength of the Al-xGnP composites continuously reduced with the addition of xGnP due to the formation of Al{sub 4}C{sub 3} particles at the interface of the Al and xGnP in the composite.

  4. The effect of graphite precipitates in Ni3Al/C composite on tribological properties

    Directory of Open Access Journals (Sweden)

    A. Janas

    2010-01-01

    Full Text Available The study shows the results of investigations of the tribological properties of cast Ni3Al/C composite and compares them with the properties of pure intermetallic phases of the Ni3Al type. An inspiration to these studies was a surprising similarity observed between the microstructure of iron-carbon alloys, and specifically of different cast iron grades, and the microstructure of, absolutely different in terms of the chemical composition, nickel-aluminium alloy. Because of carbon present in the alloy, an attempt was made to determine what effect the presence of graphite (acting as a lubricant might have on the abrasive wear behaviour of alloy during lubricated friction test. Tests were made on a Miller apparatus, used for active testing of the abrasive wear resistance. The specimen loss of mass was determined in function of time.

  5. Mechanical properties of low-density polyethylene filled by graphite nanoplatelets

    International Nuclear Information System (INIS)

    Carotenuto, G; Palomba, M; De Nicola, S; Pullini, D; Horsewell, A; Hansen, T W; Nicolais, L

    2012-01-01

    The mechanical properties of GNP/LDPE nanocomposites (graphite nanoplatelets/low density polyethylene) have been investigated, in order to establish the effect of nanoscale reinforcement within the polymer matrix. Results show that the presence of the filler does not involve a change in the microscopic structure of the polymer. However, on a macroscopic scale, GNPs limit the mobility of the polymer chains, resulting in an increase in stiffness for the final composite. Orientation of GNPs within the LDPE matrix is also an important issue that affects mechanical properties and it has been evaluated by testing nanocomposites made by different manufacturing techniques (compression moulding and blown extrusion). The comparison between the experimental data and the Halpin–Tsai model shows that the orientation of GNPs due to the extrusion process leads to values of tensile modulus higher than that obtained with the randomly oriented disposition resulting from the compression moulding technique. (paper)

  6. Spectral properties of nuclear matter

    International Nuclear Information System (INIS)

    Bozek, P

    2006-01-01

    We review self-consistent spectral methods for nuclear matter calculations. The in-medium T-matrix approach is conserving and thermodynamically consistent. It gives both the global and the single-particle properties the system. The T-matrix approximation allows to address the pairing phenomenon in cold nuclear matter. A generalization of nuclear matter calculations to the super.uid phase is discussed and numerical results are presented for this case. The linear response of a correlated system going beyond the Hartree-Fock+ Random-Phase-Approximation (RPA) scheme is studied. The polarization is obtained by solving a consistent Bethe-Salpeter (BS) equation for the coupling of dressed nucleons to an external field. We find that multipair contributions are important for the spin(isospin) response when the interaction is spin(isospin) dependent

  7. AGC-2 Graphite Preirradiation Data Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-08-01

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

  8. 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.

  9. 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.

  10. New insights into pre-lithiation kinetics of graphite anodes via nuclear magnetic resonance spectroscopy

    Science.gov (United States)

    Holtstiege, Florian; Schmuch, Richard; Winter, Martin; Brunklaus, Gunther; Placke, Tobias

    2018-02-01

    Pre-lithiation of anode materials can be an effective method to compensate active lithium loss which mainly occurs in the first few cycles of a lithium ion battery (LIB), due to electrolyte decomposition and solid electrolyte interphase (SEI) formation at the surface of the anode. There are many different pre-lithiation methods, whereas pre-lithiation using metallic lithium constitutes the most convenient and widely utilized lab procedure in literature. In this work, for the first time, solid state nuclear magnetic resonance spectroscopy (NMR) is applied to monitor the reaction kinetics of the pre-lithiation process of graphite with lithium. Based on static 7Li NMR, we can directly observe both the dissolution of lithium metal and parallel formation of LiCx species in the obtained NMR spectra with time. It is also shown that the degree of pre-lithiation as well as distribution of lithium metal on the electrode surface have a strong impact on the reaction kinetics of the pre-lithiation process and on the remaining amount of lithium metal. Overall, our findings are highly important for further optimization of pre-lithiation methods for LIB anode materials, both in terms of optimized pre-lithiation time and appropriate amounts of lithium metal.

  11. The Impact of Alkaliphilic Biofilm Formation on the Release and Retention of Carbon Isotopes from Nuclear Reactor Graphite.

    Science.gov (United States)

    Rout, S P; Payne, L; Walker, S; Scott, T; Heard, P; Eccles, H; Bond, G; Shah, P; Bills, P; Jackson, B R; Boxall, S A; Laws, A P; Charles, C; Williams, S J; Humphreys, P N

    2018-03-13

    14 C is an important consideration within safety assessments for proposed geological disposal facilities for radioactive wastes, since it is capable of re-entering the biosphere through the generation of 14 C bearing gases. The irradiation of graphite moderators in the UK gas-cooled nuclear power stations has led to the generation of a significant volume of 14 C-containing intermediate level wastes. Some of this 14 C is present as a carbonaceous deposit on channel wall surfaces. Within this study, the potential of biofilm growth upon irradiated and 13 C doped graphite at alkaline pH was investigated. Complex biofilms were established on both active and simulant samples. High throughput sequencing showed the biofilms to be dominated by Alcaligenes sp at pH 9.5 and Dietzia sp at pH 11.0. Surface characterisation revealed that the biofilms were limited to growth upon the graphite surface with no penetration of the deeper porosity. Biofilm formation resulted in the generation of a low porosity surface layer without the removal or modification of the surface deposits or the release of the associated 14 C/ 13 C. Our results indicated that biofilm formation upon irradiated graphite is likely to occur at the pH values studied, without any additional release of the associated 14 C.

  12. 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

  13. 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

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

    International Nuclear Information System (INIS)

    1996-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

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

  16. Probing the structure and mechanical properties of the graphite nodules in ductile cast irons via nano-indentation

    DEFF Research Database (Denmark)

    Andriollo, Tito; Fæster, Søren; Winther, Grethe

    2018-01-01

    Little is known today about the mechanical properties of the graphite nodules, despite the key influence these particles have on the performance of ductile cast irons. To address this issue, nano-indentation tests were performed on the cross-section of a nodule whose sub-surface morphology...

  17. Graphite for fusion energy applications

    International Nuclear Information System (INIS)

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

    1987-03-01

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

  18. Rheological and electrical properties of hybrid nanocomposites of epoxy resins filled with graphite nanoplatelets and carbon black.

    Science.gov (United States)

    Truong, Quang-Trung; Lee, Seon-Suk; Lee, Dai-Soo

    2011-02-01

    Graphite nanoplatelets (GNP) were prepared by microwave irradiation of natural graphites intercalated with ferric chloride in nitromethane (GIC). Intercalated structure of GIC was confirmed by X-ray diffraction patterns. SEM images of GIC after microwave irradiation showed the exfoliation of GIC, the formation of GNPs. Hybrid nanocomposites of bisphenol-A type epoxy resins filled with GNP and a conductive carbon black (CB) were prepared and rheological and electrical properties of the nanocomposites were investigated. Viscosity and electrical surface resistivity of the nanocomposites showed minima at certain mixtures of GNP and CB in the epoxy resins.

  19. Oxidation of PCEA nuclear graphite by low water concentrations in helium

    Energy Technology Data Exchange (ETDEWEB)

    Contescu, Cristian I., E-mail: ContescuCI@ornl.gov [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6087 (United States); Mee, Robert W. [Department of Business Analytics and Statistics, University of Tennessee, Knoxville, TN 37996-0525 (United States); Wang, Peng [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6087 (United States); Romanova, Anna V.; Burchell, Timothy D. [Department of Business Analytics and Statistics, University of Tennessee, Knoxville, TN 37996-0525 (United States)

    2014-10-15

    Accelerated oxidation tests were performed to determine kinetic parameters of the chronic oxidation reaction (i.e. slow, continuous, and persistent) of PCEA graphite in contact with helium coolant containing low moisture concentrations in high temperature gas-cooled reactors. To the authors’ knowledge such a study has not been done since the detailed analysis of reaction of H-451 graphite with steam (Velasquez, Hightower, Burnette, 1978). Since that H-451 graphite is now unavailable, it is urgently needed to characterize chronic oxidation behavior of new graphite grades that are being considered for use in gas-cooled reactors. The Langmuir–Hinshelwood mechanism of carbon oxidation by water results in a non-linear reaction rate expression, with at least six different parameters. They were determined in accelerated oxidation experiments that covered a large range of temperatures (800–1100 °C), and partial pressures of water (15–850 Pa) and hydrogen (30–150 Pa) and used graphite specimens thin enough (4 mm) in order to avoid diffusion effects. Data analysis employed a statistical method based on multiple likelihood estimation of parameters and simultaneous fitting of non-linear equations. The results show significant material-specific differences between graphite grades PCEA and H-451 which were attributed to microstructural dissimilarity between the two materials. It is concluded that kinetic data cannot be transferred from one graphite grade to another.

  20. A safety assessment of the use of graphite in nuclear reactors licensed by the US NRC

    International Nuclear Information System (INIS)

    Schweitzer, D.G.; Gurinsky, D.H.; Kaplan, E.; Sastre, C.

    1987-09-01

    This report reviews existing literature and knowledge on graphite burning and on stored energy accumulation and releases in order to assess what role, if any, a stored energy release can have in initiating or contributing to hypothetical graphite burning scenarios in research reactors. It also addresses the question of graphite ignition and self-sustained combustion in the event of a loss-of-coolant accident (LOCA). The conditions necessary to initiate and maintain graphite burning are summarized and discussed. From analyses of existing information it is concluded that only stored energy accumulations and releases below the burning temperature (650 0 C) are pertinent. After reviewing the existing knowledge on stored energy it is possible to show that stored energy releases do not occur spontaneously, and that the maximum stored energy that can be released from any reactor containing graphite is a very small fraction of the energy produced during the first few minutes of a burning incident. The conclusions from these analyses are that the potential to initiate or maintain a graphite burning incident is essentially independent of the stored energy in the graphite, and depends on other factors that are unique for these reactors, research reactors, and for Fort St. Vrain. In order to have self-sustained rapid graphite oxidation in any of these reactors, certain necessary conditions of geometry, temperature, oxygen supply, reaction product removal, and a favorable heat balance must be maintained. There is no new evidence associated with either the Windscale Accident or the Chernobyl Accident that indicates a credible potential for a graphite burning accident in any of the reactors considered in this review

  1. Effect of carbon black on electrical and rheological properties of graphite nanoplatelets/poly(ethylene-butyl acrylate composites

    Directory of Open Access Journals (Sweden)

    H. Oxfall

    2015-01-01

    Full Text Available The effect of adding carbon black on the electrical and rheological properties of graphite nanoplatelets/poly(ethylene-butyl acrylate copolymer composites produced via melt or solution mixing was studied. By adding a small amount of low- or high-structured carbon black to the nanocomposite, the electrical percolation threshold decreased and the final conductivity (at higher filler contents increased. The effect on the percolation threshold was significantly stronger in case of the high-structured carbon black where replacing 10 wt% of the total filler content with carbon black instead of graphite nanoplatelets reduced the electrical percolation threshold from 6.9 to 4.6 vol%. Finally, the solution mixing process was found to be more efficient leading to a lower percolation threshold. For the composites containing high-structured carbon black, graphite nanoplatelets and their hybrids there was a quite reasonable correlation between the electrical and rheological percolation thresholds.

  2. Influence of cooling rate and antimony addition content on graphite morphology and mechanical properties of a ductile iron

    Directory of Open Access Journals (Sweden)

    Liu Zhe

    2012-05-01

    Full Text Available Cooling rate and inoculation practice can greatly affect the graphite morphology of ductile irons. In the present research, the effects of the cooling rate and antimony addition on the graphite morphology and mechanical properties of ductile irons have been studied. Three ductile iron castings were prepared through solidification under cooling conditions S (slow, M (medium and F (fast. The cooling rates around the equilibrium eutectic temperature (1,150 ℃ for these cooling conditions (S, M and F were set at 0.21 ℃·min-1, 0.32 ℃·min-1 and 0.37 ℃·min-1, respectively. In addition, four ductile iron castings were prepared by adding 0.01%, 0.02%, 0.03% and 0.04% (by weight antimony, respectively under the slow cooling condition. The results show that the nodularity index, tensile strength and hardness of the ductile iron castings without antimony addition are all improved with the increase of cooling rate, while the ductile iron casting solidified under the medium cooling rate possesses the largest number of graphite nodules. Furthermore, for the four antimony containing castings, the graphite morphology and tensile strength are also improved by the antimony additions, and the effect of antimony addition is intensified when the addition increases from 0.01% to 0.03%. Moreover, the rare earth elements (REE/antimony ratio of 2 appears to be the most effective for fine nodular graphite formation in ductile iron.

  3. Electronic structure and physical properties of 13C carbon composite

    OpenAIRE

    Zhmurikov, Evgenij

    2015-01-01

    This review is devoted to the application of graphite and graphite composites in science and technology. Structure and electrical properties, as so technological aspects of producing of high-strength artificial graphite and dynamics of its destruction are considered. These type of graphite are traditionally used in the nuclear industry. Author was focused on the properties of graphite composites based on carbon isotope 13C. Generally, the review relies on the original results and concentrates...

  4. Morphological and electrical properties of epoxy-based composites reinforced with exfoliated graphite

    Energy Technology Data Exchange (ETDEWEB)

    Lamberti, Patrizia; Spinelli, Giovanni, E-mail: gspinelli@unisa.it; Tucci, Vincenzo [Department of Information and Electrical Engineering and Applied Mathematics University of Salerno, Via Giovanni Paolo II, Fisciano (Italy); Guadagno, Liberata; Raimondo, Marialuigia; Vertuccio, Luigi [Department of Industrial Engineering University of Salerno, Via Giovanni Paolo II, Fisciano (Italy)

    2016-05-18

    An experimental study has been carried out to prepare and characterize epoxy/amine-based composites filled with different percentages of partially exfoliated graphite (i.e. pEG) particles having an exfoliation degree of 56% in order to analyze the effect of the filler amounts on the electrical properties of the resulting nanocomposites. Moreover, in order to fully investigate the direct relationship between the physical properties of the employed filler and the results of the electrical characterization, a structural and morphological characterization of the pEG samples is carried out by means of various type of analysis such as X-ray diffraction patterns, micro-Raman and Scanning Electron Microscopy (SEM) images. The DC electrical characterization reveals a percolation thresholds (EPT) that falls in the range [2–3] wt% and an electrical conductivity of about 0.66 S/m at the highest filler loading (6.5 wt%). From the analysis of the percolative curve it is possible to derive the percolation law parameters and in particular the critical exponent t, whose value (i.e. 1.2) reflects an effective 2D organization of the percolating structure consistent with the type of filler used (2-dimensional). Finally, an extensive analysis concerning the electrical properties in the frequency domain has been carried out in order to evaluate the effectiveness of pEG-loaded composites in terms of electromagnetic interference compatibility (EMC) and their applicability as radar absorbers materials (RAMs).

  5. Evaluation of microstructures and oxidation behaviors of graphite for core support structure

    International Nuclear Information System (INIS)

    Park, Soo Jin; Bae, Kyung Min

    2010-03-01

    This work studies the oxidation-induced characteristics of five nuclear graphites (NBG-17, NBG-18, NBG-25, IG-110, and IG-430). The oxidation characteristics of the nuclear graphites were measured at 600 .deg. C. The surface properties of the oxidation graphites were characterized by means of scanning electron microscopy, X-ray photoelectron spectroscopy, and contact angle methods. The N2/77K adsorption isotherm characteristics, including the specific surface area and micropore volume, were investigated by means of BET and t-plot methods. The experimental results show an increase in the average pore size of graphites; they also show that oxidation produces the surface functional groups on the graphite surfaces. The surface area of each graphite behaves in a unique manner. For example the surface area of NBG-17 increases slightly whereas the surface area of IG-110 increases significantly. This result confirms that the original surface state of each graphite is unique

  6. Optimisation of the link volume for weakest link failure prediction in NBG-18 nuclear graphite

    International Nuclear Information System (INIS)

    Hindley, Michael P.; Groenwold, Albert A.; Blaine, Deborah C.; Becker, Thorsten H.

    2014-01-01

    This paper describes the process for approximating the optimal size of a link volume required for weakest link failure calculation in nuclear graphite, with NBG-18 used as an example. As part of the failure methodology, the link volume is defined in terms of two grouping criteria. The first criterion is a factor of the maximum grain size and the second criterion is a function of an equivalent stress limit. A methodology for approximating these grouping criteria is presented. The failure methodology employs finite element analysis (FEA) in order to predict the failure load, at 50% probability of failure. The average experimental failure load, as determined for 26 test geometries, is used to evaluate the accuracy of the weakest link failure calculations. The influence of the two grouping criteria on the failure load prediction is evaluated by defining an error in prediction across all test cases. Mathematical optimisation is used to find the minimum error across a range of test case failure predictions. This minimum error is shown to deliver the most accurate failure prediction across a whole range of components, although some test cases in the range predict conservative failure load. The mathematical optimisation objective function is penalised to account for non-conservative prediction of the failure load for any test case. The optimisation is repeated and a link volume found for conservative failure prediction. The failure prediction for each test case is evaluated, in detail, for the proposed link volumes. Based on the analysis, link design volumes for NBG-18 are recommended for either accurate or conservative failure prediction

  7. Preparation and tribological properties of self-lubricating TiO2/graphite composite coating on Ti6Al4V alloy

    International Nuclear Information System (INIS)

    Mu, Ming; Zhou, Xinjian; Xiao, Qian; Liang, Jun; Huo, Xiaodi

    2012-01-01

    Highlights: ► A TiO 2 /graphite composite coating is produced on Ti alloy by one-step PEO process. ► The TiO 2 /graphite composite coating exhibits excellent self-lubricating behavior. ► The self-lubricating composite coating improves the wear resistance by comparison to the conventional PEO coating. - Abstract: One-step plasma electrolytic oxidation (PEO) process in a graphite-dispersed phosphate electrolyte was used to prepare a graphite-containing oxide composite coating on Ti6Al4V alloy. The composition and microstructure of the oxide coatings produced in the phosphate electrolytes with and without addition of graphite were analyzed by X-ray diffractometer (XRD), Raman spectroscopy, X-ray photoelectron spectroscopy (XPS) and scanning electron microscopy (SEM). The tribological properties of the uncoated Ti6Al4V alloy and oxide coatings were evaluated using a reciprocating ball-on-disk tribometer. Results showed that the graphite-containing oxide composite coating can be successfully produced on Ti6Al4V alloy in the graphite-dispersed phosphate electrolyte using PEO process. The graphite-containing oxide composite coating registered much lower friction coefficient and wear rate than the uncoated Ti6Al4V alloy and the oxide coating without graphite under dry sliding condition, exhibiting excellent self-lubricating property.

  8. An Facile High-Density Polyethylene - Exfoliated Graphite - Aluminium Hydroxide Composite: Manufacture, Morphology, Structure, Antistatic and Fireproof Properties

    Directory of Open Access Journals (Sweden)

    Jihui LI

    2014-09-01

    Full Text Available Graphite intercalation compounds (GIC and exfoliated graphite (EG as raw materials were prepared with flake graphite, concentrated sulphuric acid (H2SO4, potassium bichromate (K2Cr2O7 and peracetic acid (CH3CO3H and characterized. Then, high-density polyethylene-exfoliated graphite (HDPE-EG composites were fabricated with HDPE and EG via in situ synthesis technique in the different mass ratio, and their resistivity values (ohms/sq were measured. Based on the resistivity values, it was discovered that HDPE-EG composite with the antistatic property could be fabricated while the mass ratio was 5.00 : 0.30. Last, HDPE-EG-aluminium hydroxide (HDPE-EG-Al(OH3 composites were manufactured with HDPE, GIC and Al(OH3 via the in situ synthesis-thermal expansion technique, and their resistivity values and limiting oxygen index (LOI values were measured. Based on the resistivity values and LOI values, it was discovered that HDPE-EG-Al(OH3 composite with the antistatic and fireproof property could be manufactured while HDPE, GICs and Al(OH3 of mass ratio was 5.00 : 0.30 : 1.00. Otherwise, the petal-like morphology and structure of HDPE-EG-Al(OH3 composite were characterized, which consisted of EG, HDPE and Al(OH3. DOI: http://dx.doi.org/10.5755/j01.ms.20.3.4275

  9. Plasmon mediated enhancement and tuning of optical emission properties of two dimensional graphitic carbon nitride nanosheets.

    Science.gov (United States)

    Bayan, Sayan; Gogurla, Narendar; Midya, Anupam; Singha, Achintya; Ray, Samit K

    2017-12-01

    We demonstrate surface plasmon induced enhancement and tunablilty in optical emission properties of two dimensional graphitic carbon nitride (g-C 3 N 4 ) nanosheets through the attachment of gold (Au) nanoparticles. Raman spectroscopy has revealed surface enhanced Raman scattering that arises due to the combined effect of the charge transfer process and localized surface plasmon induced enhancement in electromagnetic field, both occurring at the nanoparticle-nanosheet interface. Photoluminescence studies suggest that at an optimal concentration of nanoparticles, the emission intensity can be enhanced, which is maximum within the 500-525 nm region. Further, the fabricated electroluminescent devices reveal that the emission feature can be tuned from bluish-green to red (∼160 nm shift) upon attaching Au nanoparticles. We propose that the π*→π transition in g-C 3 N 4 can trigger surface plasmon oscillation in Au, which subsequently increases the excitation process in the nanosheets and results in enhanced emission in the green region of the photoluminescence spectrum. On the other hand, electroluminescence of g-C 3 N 4 can induce plasmon oscillation more efficiently and thus can lead to red emission from Au nanoparticles through the radiative damping of particle plasmons. The influence of nanoparticle size and coverage on the emission properties of two dimensional g-C 3 N 4 , nanosheets has also been studied in detail.

  10. Microwave absorbing property of silicone rubber composites with added carbonyl iron particles and graphite platelet

    International Nuclear Information System (INIS)

    Xu, Yonggang; Zhang, Deyuan; Cai, Jun; Yuan, Liming; Zhang, Wenqiang

    2013-01-01

    Silicone rubber composites filled with carbonyl iron particles (CIPs) and graphite platelet (GP) were prepared using non-coating or coating processes. The complex permittivity and permeability of the composites were measured using a vector network analyzer in the frequency range of 1–18 GHz and dc electric conductivity was measured by the standard four-point contact method. The results showed that CIPs/GP composites fabricated in the coating process had the highest permittivity and permeability due to the particle orientation and interactions between the two absorbents. The coating process resulted in a decreased effective eccentricity of the absorbents, and the dc conductivity increased according to Neelakanta's equations. The reflection loss (RL) value showed that the composites had an excellent absorbing property in the L-band, minimum −11.85 dB at 1.5 mm and −15.02 dB at 2 mm. Thus, GP could be an effective additive in preparing thin absorbing composites in the L-band. - Highlights: ► The added GP increased the permittivity and permeability of composites filled with CIPs. ► The enhancement was owing to interactions of the two absorbents and the fabrication process. ► The coating process decreased the effective eccentricity of the particles, and increased the conductivity of the composites. ► The composites to which CIPs/GP were added in coating process had excellent absorbing properties in the L-band.

  11. 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.

  12. Titanium Powder Sintering in a Graphite Furnace and Mechanical Properties of Sintered Parts

    Directory of Open Access Journals (Sweden)

    Changzhou Yu

    2017-02-01

    Full Text Available Recent accreditation of titanium powder products for commercial aircraft applications marks a milestone in titanium powder metallurgy. Currently, powder metallurgical titanium production primarily relies on vacuum sintering. This work reported on the feasibility of powder sintering in a non-vacuum furnace and the tensile properties of the as-sintered Ti. Specifically, we investigated atmospheric sintering of commercially pure (C.P. titanium in a graphite furnace backfilled with argon and studied the effects of common contaminants (C, O, N on sintering densification of titanium. It is found that on the surface of the as-sintered titanium, a severely contaminated porous scale was formed and identified as titanium oxycarbonitride. Despite the porous surface, the sintered density in the sample interiors increased with increasing sintering temperature and holding time. Tensile specimens cut from different positions within a large sintered cylinder reveal different tensile properties, strongly dependent on the impurity level mainly carbon and oxygen. Depending on where the specimen is taken from the sintered compact, ultimate tensile strength varied from 300 to 580 MPa. An average tensile elongation of 5% to 7% was observed. Largely depending on the interstitial contents, the fracture modes from typical brittle intergranular fracture to typical ductile fracture.

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  14. Gross properties of nuclei and nuclear excitations

    International Nuclear Information System (INIS)

    Feldmeier, H.

    1990-01-01

    The workshop contains discussions about theory and experiments with heavy ions at intermediate and relativistic energies, spallation, quark matter production and other nuclear matter properties. (BBOE)

  15. Mechanical and physical properties of carbon-graphite fiber-reinforced polymers intended for implant suprastructures.

    Science.gov (United States)

    Segerström, Susanna; Ruyter, I Eystein

    2007-09-01

    Mechanical properties and quality of fiber/matrix adhesion of poly(methyl methacrylate) (PMMA)-based materials, reinforced with carbon-graphite (CG) fibers that are able to remain in a plastic state until polymerization, were examined. Tubes of cleaned braided CG fibers were treated with a sizing resin. Two resin mixtures, resin A and resin B, stable in the fluid state and containing different cross-linking agents, were reinforced with CG fiber loadings of 24, 36, and 47 wt% (20, 29, and 38 vol.%). In addition, resin B was reinforced with 58 wt% (47 vol.%). After heat-polymerization, flexural strength and modulus were evaluated, both dry and after water storage. Coefficient of thermal expansion, longitudinally and in the transverse direction of the specimens, was determined. Adhesion between fibers and matrix was evaluated with scanning electron microscopy (SEM). Flexural properties and linear coefficient of thermal expansion were similar for both fiber composites. With increased fiber loading, flexural properties increased. For 47 wt% fibers in polymer A the flexural strength was 547.7 (28.12) MPa and for polymer B 563.3 (89.24) MPa when water saturated. Linear coefficient of thermal expansion was for 47 wt% CG fiber-reinforced polymers; -2.5 x 10(-6) degrees C-1 longitudinally and 62.4 x 10(-6) degrees C-1 in the transverse direction of the specimens. SEM revealed good adhesion between fibers and matrix. More porosity was observed with fiber loading of 58 wt%. The fiber treatment and the developed resin matrices resulted in good adhesion between CG fibers and matrix. The properties observed indicate a potential for implant-retained prostheses.

  16. 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

  17. Correlation between Mechanical Properties with Specific Wear Rate and the Coefficient of Friction of Graphite/Epoxy Composites

    Directory of Open Access Journals (Sweden)

    Mahdi Alajmi

    2015-07-01

    Full Text Available The correlation between the mechanical properties of Fillers/Epoxy composites and their tribological behavior was investigated. Tensile, hardness, wear, and friction tests were conducted for Neat Epoxy (NE, Graphite/Epoxy composites (GE, and Data Palm Fiber/Epoxy with or without Graphite composites (GFE and FE. The correlation was made between the tensile strength, the modulus of elasticity, elongation at the break, and the hardness, as an individual or a combined factor, with the specific wear rate (SWR and coefficient of friction (COF of composites. In general, graphite as an additive to polymeric composite has had an eclectic effect on mechanical properties, whereas it has led to a positive effect on tribological properties, whilst date palm fibers (DPFs, as reinforcement for polymeric composite, promoted a mechanical performance with a slight improvement to the tribological performance. Statistically, this study reveals that there is no strong confirmation of any marked correlation between the mechanical and the specific wear rate of filler/Epoxy composites. There is, however, a remarkable correlation between the mechanical properties and the friction coefficient of filler/Epoxy composites.

  18. Correlation between Mechanical Properties with Specific Wear Rate and the Coefficient of Friction of Graphite/Epoxy Composites.

    Science.gov (United States)

    Alajmi, Mahdi; Shalwan, Abdullah

    2015-07-08

    The correlation between the mechanical properties of Fillers/Epoxy composites and their tribological behavior was investigated. Tensile, hardness, wear, and friction tests were conducted for Neat Epoxy (NE), Graphite/Epoxy composites (GE), and Data Palm Fiber/Epoxy with or without Graphite composites (GFE and FE). The correlation was made between the tensile strength, the modulus of elasticity, elongation at the break, and the hardness, as an individual or a combined factor, with the specific wear rate (SWR) and coefficient of friction (COF) of composites. In general, graphite as an additive to polymeric composite has had an eclectic effect on mechanical properties, whereas it has led to a positive effect on tribological properties, whilst date palm fibers (DPFs), as reinforcement for polymeric composite, promoted a mechanical performance with a slight improvement to the tribological performance. Statistically, this study reveals that there is no strong confirmation of any marked correlation between the mechanical and the specific wear rate of filler/Epoxy composites. There is, however, a remarkable correlation between the mechanical properties and the friction coefficient of filler/Epoxy composites.

  19. Property Analysis of Exfoliated Graphite Nanoplatelets Modified Asphalt Model Using Molecular Dynamics (MD Method

    Directory of Open Access Journals (Sweden)

    Hui Yao

    2017-01-01

    Full Text Available This Molecular Dynamics (MD simulation paper presents a physical property comparison study between exfoliated graphite nanoplatelets (xGNP modified and control asphalt models, including density, glass transition temperature, viscosity and thermal conductivity. The three-component control asphalt model consists of asphaltenes, aromatics, and saturates based on previous references. The xGNP asphalt model was built by incorporating an xGNP and control asphalt model and controlling mass ratios to represent the laboratory prepared samples. The Amber Cornell Extension Force Field (ACEFF was used with assigned molecular electro-static potential (ESP charge from NWChem analysis. After optimization and ensemble relaxation, the properties of the control and xGNP modified asphalt models were computed and analyzed using the MD method. The MD simulated results have a similar trend as the test results. The property analysis showed that: (1 the density of the xGNP modified model is higher than that of the control model; (2 the glass transition temperature of the xGNP modified model is closer to the laboratory data of the Strategic Highway Research Program (SHRP asphalt binders than that of the control model; (3 the viscosities of the xGNP modified model at different temperatures are higher than those of the control model, and it coincides with the trend in the laboratory data; (4 the thermal conductivities of the xGNP modified asphalt model are higher than those of the control asphalt model at different temperatures, and it is consistent with the trend in the laboratory data.

  20. Further investigation on boric acid catalytic graphitization of polyacrylonitrile carbon fibers: Mechanism and mechanical properties

    International Nuclear Information System (INIS)

    Wen, Ya; Lu, Yonggen; Xiao, Hao; Qin, Xianying

    2012-01-01

    Highlights: ► The modulus of carbon fiber was improved by boric acid at the temperature range of 1500–2900 °C. ► 2300 °C is a key temperature degree from which the boron began to benefit fiber strength. ► The fiber strength was affected by the boron reaction and related to the boron states. -- Abstract: Catalytic graphitization of polyacrylonitrile based carbon fibers by boric acid doping was studied and the dependence of fiber tensile strength on the boron content and temperature was discussed. It was found that there existed a key temperature point for the boron to take effect. When the fibers were modified with 7.0 wt.% boric acid solution, with increasing temperature, the tensile strength was lower than that of the unmodified ones below 2300 °C, but a reverse thing happened above 2300 °C. Moreover, when being heated at 2500 °C, the modified fibers showed an increasing tensile modulus and strength with increasing boron content till maximums of 404 GPa and 2.46 GPa, 26% and 16% higher than those of unmodified ones. The mechanical properties of the fibers were affected by the interaction of carbon and boron, and also related with boron states. The decomposition of boron acid and its interaction with carbon brought defects on fiber surface, degrading the mechanical properties below 1300 °C. With further heat treatment, the boron diffused into the fibers and divided into two states: substitutional and interstitial. At a temperature over 2300 °C with an appreciate boron content, the substitutional would be formed predominantly, which removed the structural defects and relaxed the distortions, so as to benefit the mechanical properties.

  1. 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

  2. The Effect of Hydrogen on the Mechanical Properties of Cast Irons and ADI with Various Carbon Equivalent and Graphite Morphology

    International Nuclear Information System (INIS)

    Cho, Yong Gi; Lee, Kyung Sub

    1989-01-01

    The effect of hydrogen on the mechanical properties of cast irons, flake, CV graphite cast iron ductile iron and ADI have been investigated. The effects of various carbon equivalent, graphite morphology and matrix have been analyzed to determine the predominant factor which influences on the hydrogen embrittlement. The effect of various carbon equivalent on the embrittlement was little in the similar graphite morphology. The embrittlement of ferrite matrix changed by heat treatment was less than that of pearlite matrix. In the case of ADI, the tendency of hydrogen embrittlement of lower bainite matrix was less remarkable than that of upper banite matrix. As the result of hydrogen charging, the tendency of interface decohesion between matrix-graphite was increased in flake G.C.I., and the trend from ductile fracture mode to brittle fracture mode was observed in CV G.C.I and ductile iron. Lower bainite in ADI showed the ductile fracture mode. Hydrogen solubility of lower bainite was higher than that of upper bainite

  3. GEO-MIX-SELF calculations of the elastic properties of a textured graphite sample at different hydrostatic pressures

    International Nuclear Information System (INIS)

    Matthies, Siegfried

    2012-01-01

    The recently developed GEO-MIX-SELF approximation (GMS) is applied to interpret the pressure dependence of the longitudinal ultrasonic wave velocities in a polycrystalline graphite sample that has already been investigated in a wide range of experimental contexts. Graphite single crystals have extremely anisotropic elastic properties, making this sample a challenging test to demonstrate the potential of the GMS method. GMS combines elements of well known self-consistent algorithms and of the geometric mean approximation. It is able to consider mixtures of different polycrystalline phases, each with its own nonspherical grain shape and preferred orientation (texture). Pores and 'cracks', typical for bulk graphite, are modeled as phases with 'empty' grains. The pressure dependence (up to 150 MPa) of the experimental wave velocities can be well explained using the known texture of the sample by fitting the shape parameters and volume fractions of the graphite grains, cracks and spherical pores. The pressure dependence of these parameters describes a reasonable scenario for the closing of the cracks and pores with increasing pressure. (orig.)

  4. Oxidation Resistant Graphite Studies

    Energy Technology Data Exchange (ETDEWEB)

    W. Windes; R. Smith

    2014-07-01

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

  5. Defect induced tuning of photoluminescence property in graphitic carbon nitride nanosheets through synthesis conditions

    Energy Technology Data Exchange (ETDEWEB)

    Das, D. [School of Materials Science and Nanotechnology, Jadavpur University, Kolkata 700032 (India); Banerjee, D., E-mail: nilju82@gmail.com [School of Materials Science Engineering Indian Institute of Engineering Science and Technology, Shibpur, Howrah (India); Pahari, D. [School of Materials Science Engineering Indian Institute of Engineering Science and Technology, Shibpur, Howrah (India); Ghorai, U.K. [Department of Industrial Chemistry & Swami Vivekananda Research centre, Ramakrishna Mission Vidyamandira, Belur Math, Howrah 711202 (India); Sarkar, S.; Das, N.S. [School of Materials Science and Nanotechnology, Jadavpur University, Kolkata 700032 (India); Chattopadhyay, K.K., E-mail: kalyan_chattopadhyay@yahoo.com [School of Materials Science and Nanotechnology, Jadavpur University, Kolkata 700032 (India); Thin Film and Nanoscience Laboratory, Department of Physics, Jadavpur University, Kolkata 700032 (India)

    2017-05-15

    Synthesis of layered sheet like graphitic carbon nitride by pyrolysis of urea at different temperatures has been reported. The proper phase formation has been confirmed by X-ray diffraction study whereas field emission scanning and transmission electron microscope characterized the morphology of the material. Fourier transform infrared and Raman spectroscopy revealed the presence of different bonds in the sample. Thermal gravimetric analysis has been used to study the thermal stability of the material. Energy dispersive X-ray analysis further revealed the elemental composition of carbon and nitrogen in a proper stoichiometric ratio. Excitation dependent photoluminescence spectra of the as prepared samples have been studied in detail. It has been shown that synthesis condition can tailor the amount of defects present in the synthesized samples that in turn can change the photoluminescence properties of the material. The fluorescence spectra of the as prepared samples have been used to detect copper ions present in the sample. It has also been shown that the presence of defects which is mainly N-H functional groups can change the decay characteristics of the carrier in these samples which in turn changes the PL spectra.

  6. Electrical Properties of Cement-Based Composites with Carbon Nanotubes, Graphene, and Graphite Nanofibers.

    Science.gov (United States)

    Yoo, Doo-Yeol; You, Ilhwan; Lee, Seung-Jung

    2017-05-08

    This study was conducted to evaluate the effect of the carbon-based nanomaterial type on the electrical properties of cement paste. Three different nanomaterials, multi-walled carbon nanotubes (MWCNTs), graphite nanofibers (GNFs), and graphene (G), were incorporated into the cement paste at a volume fraction of 1%. The self-sensing capacity of the cement composites was also investigated by comparing the compressive stress/strain behaviors by evaluating the fractional change of resistivity (FCR). The electrical resistivity of the plain cement paste was slightly reduced by adding 1 vol % GNFs and G, whereas a significant decrease of the resistivity was achieved by adding 1 vol % MWCNTs. At an identical volume fraction of 1%, the composites with MWCNTs provided the best self-sensing capacity with insignificant noise, followed by the composites containing GNFs and G. Therefore, the addition of MWCNTs was considered to be the most effective to improve the self-sensing capacity of the cement paste. Finally, the composites with 1 vol % MWCNTs exhibited a gauge factor of 113.2, which is much higher than commercially available strain gauges.

  7. Electrical Properties of Cement-Based Composites with Carbon Nanotubes, Graphene, and Graphite Nanofibers

    Directory of Open Access Journals (Sweden)

    Doo-Yeol Yoo

    2017-05-01

    Full Text Available This study was conducted to evaluate the effect of the carbon-based nanomaterial type on the electrical properties of cement paste. Three different nanomaterials, multi-walled carbon nanotubes (MWCNTs, graphite nanofibers (GNFs, and graphene (G, were incorporated into the cement paste at a volume fraction of 1%. The self-sensing capacity of the cement composites was also investigated by comparing the compressive stress/strain behaviors by evaluating the fractional change of resistivity (FCR. The electrical resistivity of the plain cement paste was slightly reduced by adding 1 vol % GNFs and G, whereas a significant decrease of the resistivity was achieved by adding 1 vol % MWCNTs. At an identical volume fraction of 1%, the composites with MWCNTs provided the best self-sensing capacity with insignificant noise, followed by the composites containing GNFs and G. Therefore, the addition of MWCNTs was considered to be the most effective to improve the self-sensing capacity of the cement paste. Finally, the composites with 1 vol % MWCNTs exhibited a gauge factor of 113.2, which is much higher than commercially available strain gauges.

  8. Process industry properties in nuclear industry

    International Nuclear Information System (INIS)

    Zheng Hualing

    2005-01-01

    In this article the writer has described the definition of process industry, expounded the fact classifying nuclear industry as process industry, compared the differences between process industry and discrete industry, analysed process industry properties in nuclear industry and their important impact, and proposed enhancing research work on regularity of process industry in nuclear industry. (authors)

  9. The determination of the elastic properties of an anisotropic polycrystalline graphite using neutron diffraction and ultrasonic measurements

    Czech Academy of Sciences Publication Activity Database

    Lokajíček, Tomáš; Lukáš, Petr; Nikitin, A. N.; Papushkin, I.V.; Sumin, V. V.; Vasin, R.N.

    2010-01-01

    Roč. 49, č. 4 (2010), s. 1374-1384 ISSN 0008-6223 R&D Projects: GA ČR GA205/08/0676 Institutional research plan: CEZ:AV0Z30130516; CEZ:AV0Z10480505 Keywords : extruded graphite * elastic properties * neutron diffraction * ultrasonic sounding * thermal-expansion * self-consistent * young moduls * porosity * stress * rocks Subject RIV: DB - Geology ; Mineralogy Impact factor: 4.893, year: 2010

  10. The correlation of low-velocity impact resistance of graphite-fiber-reinforced composites with matrix properties

    Science.gov (United States)

    Bowles, Kenneth J.

    1988-01-01

    Summarized are basic studies that were conducted to correlate the impact resistance of graphite-fiber-reinforced composites with polymer matrix properties. Three crosslinked epoxy resins and a linear polysulfone were selected as composite matrices. As a group, these resins possess a significantly large range of mechanical properties. The mechanical properties of the resins and their respective composites were measured. Neat resin specimens and unidirectional and crossply composite specimens were impact tested with an instrumented dropweight tester. Impact resistances of the specimens were assesseed on the basis of loading capability, energy absorption, and extent of damage.

  11. Electromagnetic properties and performance of exfoliated graphite (EG) - thermoplastic polyurethane (TPU) nanocomposites at microwaves

    Czech Academy of Sciences Publication Activity Database

    Valentini, M.; Piana, Francesco; Pionteck, J.; Lamastra, F. R.; Nanni, F.

    2015-01-01

    Roč. 114, 19 June (2015), s. 26-33 ISSN 0266-3538 Institutional support: RVO:61389013 Keywords : exfoliated graphite * functional composites * nano composites Subject RIV: CD - Macromolecular Chemistry Impact factor: 3.897, year: 2015

  12. A Nanotube Surface Reinforced Graphite Fiber Exhibiting Significantly Enhanced Properties, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — The completed Phase I work was directed at the application of nanotechnology to graphite/epoxy composites. A novel approach to the application of the nanotubes onto...

  13. The effect of compressive stress on the Young's modulus of unirradiated and irradiated nuclear graphites

    International Nuclear Information System (INIS)

    Oku, T.; Usui, T.; Ero, M.; Fukuda, Y.

    1977-01-01

    The Young's moduli of unirradiated and high temperature (800 to 1000 0 C) irradiated graphites for HTGR were measured by the ultrasonic method in the direction of applied compressive stress during and after stressing. The Young's moduli of all the tested graphites decreased with increasing compressive stress both during and after stressing. In order to investigate the reason for the decrease in Young's modulus by applying compressive stress, the mercury pore diameter distributions of a part of the unirradiated and irradiated specimens were measured. The change in pore distribution is believed to be associated with structural changes produced by irradiation and compressive stressing. The residual strain, after removing the compressive stress, showed a good correlation with the decrease in Young's modulus caused by the compressive stress. The decrease in Young's modulus by applying compressive stress was considered to be due to the increase in the mobile dislocation density and the growth or formation of cracks. The results suggest, however, that the mechanism giving the larger contribution depends on the brand of graphite, and in anisotropic graphite it depends on the direction of applied stress and the irradiation conditions. (author)

  14. Nuclear radiation and the properties of concrete

    International Nuclear Information System (INIS)

    Kaplan, M.F.

    1983-08-01

    Concrete is used for structures in which the concrete is exposed to nuclear radiation. Exposure to nuclear radiation may affect the properties of concrete. The report mentions the types of nuclear radiation while radiation damage in concrete is discussed. Attention is also given to the effects of neutron and gamma radiation on compressive and tensile strength of concrete. Finally radiation shielding, the attenuation of nuclear radiation and the value of concrete as a shielding material is discussed

  15. Graphite intercalated polyaniline composite with superior anticorrosive and hydrophobic properties, as protective coating material on steel surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Rathnayake, R.M.N.M. [National Institute of Fundamental Studies, Kandy (Sri Lanka); Graduate School of Engineering, Toyota Technological Institute, 2-12-1 Hisakata, Tempaku, Nagoya 468-8511 (Japan); Mantilaka, M.M.M.G.P.G. [Sri Lanka Institute of Nanotechnology, Nanotechnology and Science Park, Mahenwatte, Pitipana, Homagama (Sri Lanka); Hara, Masanori; Huang, Hsin-Hui [Graduate School of Engineering, Toyota Technological Institute, 2-12-1 Hisakata, Tempaku, Nagoya 468-8511 (Japan); Wijayasinghe, H.W.M.A.C., E-mail: athula@ifs.ac.lk [National Institute of Fundamental Studies, Kandy (Sri Lanka); Yoshimura, Masamichi [Graduate School of Engineering, Toyota Technological Institute, 2-12-1 Hisakata, Tempaku, Nagoya 468-8511 (Japan); Pitawala, H.M.T.G.A. [Department of Geology, University of Peradeniya, Peradeniya (Sri Lanka)

    2017-07-15

    Highlights: • In this paper, it has been utilized a novel method to prepare a new composite material of PANI/NPG graphite composite, using NPG vein graphite variety. • It is found that the composite works as an anti-corrosive coating on steel surfaces. Further, the prepared composite shows good hydrophobic ability, which is very useful in preventing corrosion on metal surfaces. • The prepared PANI/NPG composite material shows a significantly high corrosion resistance compared to alkyd resin/PANI coatings or alkyd resin coatings, on steel surfaces. - Abstract: Solid polymer composite systems are widely being used for potential technological applications in secondary energy sources and electrochromic devices. In this study, we synthesized and characterized a composite material composed of polyaniline (PANI) and natural needle platy (NPG) vein graphite. Fourier transform infrared spectroscopy (FTIR), X-ray diffraction (XRD), cyclic voltammetry (CV), scanning electron microscopy (SEM), X-ray photoelectron spectroscopy (XPS), micro-Raman analysis, thermogravimetric and differential thermal analysis (TGA/DTA), transmission electron microscopy (TEM) were used to study the structural and electrochemical properties of the prepared PANI/NPG graphite composite. XPS, FTIR, and micro-Raman analysis confirmed the existence of relevant functional groups and bonding in the prepared PANI/NPG composite material. The composite shows a very low corrosion rate, approximately 29 μm per year, and high hydrophobicity on steel surfaces, which helps to prevent the corrosion due to O{sub 2} penetration towards the metal surface. It indicates that the composite can be used as a high potential surface coating material to anticorrosion. The specific capacitance of PANI/NPG composite is 833.3 F g{sup −1}, which is higher than that of PANI. This synergistic electrical performance result proves the prepared PANI/NPG graphite composite as a suitable protective coating material for steel

  16. Dynamic method for the measurement of Young'S modulus. Application to nuclear graphites; Methode de mesure dynamique du module d'Young. Application aux graphites nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Pattou, F; Trutt, J C

    1963-07-01

    A dynamic method has been developed for measuring Young's modulus and the rigidity modulus using the 'Forster Elastomat'. The principle consists in the determination of the resonance frequencies of graphite samples submitted to transverse, longitudinal, and torsional vibrations. The first two modes of vibration make it possible to calculate the elasticity modulus or the Young's modulus E, the third mode makes possible the calculation of the rigidity modulus G. The relationships from which the moduli E and G are measured are given. A systematic study has been made of graphite samples produced by extrusion or compression and submitted afterwards to one or several impregnations with pitch. For graphites made from the same coke by the same method, a linear relationship has been found for Young's modulus as a function of the apparent density. For the same apparent density, graphites made from different starting materials have generally different Young's moduli that bear a relationship to the crystalline characteristics of the material. The measurements of the rigidity modulus C made on different graphites also show the influence of crystallite orientation. (authors) [French] Une methode de mesure dynamique du module d'Young et du module de rigidite du graphite utilisant 'l'Elastomat Forster' a ete mise au point. Le principe consiste a determiner les frequences de resonance d'echantillons de graphite soumis a des vibrations transversales, longitudinales et de torsion. Les deux premiers modes de vibration permettent de calculer le module d'elasticite ou module d'Young E, le troisieme mode de vibration permet de calculer le module de rigidite G. Apres avoir decrit la methode de mesure, on rappelle les relations qui permettent de calculer les modules E et G. L'etude systematique d'echantillons de graphite, fabriques par filage ou pressage et ayant subi eventuellement une ou plusieurs impregnations au brai a ete effectuee. Pour les graphites issus du meme coke et fabriques

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

    International Nuclear Information System (INIS)

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

    2006-10-01

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

  18. Nuclear property insurance: status and outlook

    International Nuclear Information System (INIS)

    1982-05-01

    The report addresses the problem of the unavailability of adequate levels of property insurance for commercial power reactors to pay for decontamination and cleanup costs arising from accidents. The report is designed to answer six questions, as follows: (1) What has been the development of each principal source of nuclear property insurance used as of early 1982 by nuclear utilities in the United States; (2) What are some of the distinguishing features of nuclear property insurance as offered by the principal sources; (3) How much nuclear property insurance was offered by each of these sources as of January 1, 1982; (4) Assuming that present plans came to fruition, how much nuclear property insurance is likely to be offered by each of these sources as of January 1, 1983; (5) What, if any, principal sources of nuclear property insurance are likely to emerge in the private sector by January 1, 1983; (6) What problems serious enough to warrant action of the NRC exist with respect to nuclear property insurance and what action should NRC take in response to each problem

  19. Microwave absorption properties of NiCoFe2O4-graphite embedded poly(o-phenetidine nanocomposites

    Directory of Open Access Journals (Sweden)

    Anil Ohlan

    2011-09-01

    Full Text Available Poly(o-phenetidine nanocomposites (PNG with NiCoFe2O4 and exfoliated graphite have been synthesized via in-situ emulsion polymerization. Systematic investigations reveal that the NiCoFe2O4 nanoparticles (30-40 nm in the poly(o-phenetidine matrix have phenomenal effect in determining the electrical, magnetic, and the microwave absorption properties of the nanocomposites. Shielding effectiveness due to absorption (SEA value of 32 dB (>99.9% has been achieved for PNG composite for its use as broadband microwave absorbing material. The microwave absorption of these composites can be attributed to dielectric loss from graphite and poly(o-phenetidine matrix, and magnetic loss from NiCoFe2O4 nanoparticles.

  20. Preparation and magnetic properties of polymer magnetic composites based on acrylate resin filled with nickel plating graphite nanosheets

    Energy Technology Data Exchange (ETDEWEB)

    Zhang Yi, E-mail: zhangyi520love@yahoo.com.cn [Department of Applied Chemistry, School of Science, Northwestern Polytechnical University, Xi' an 710072 (China); Qi Shuhua; Zhang Fan; Yang Yongqing; Duan Guochen [Department of Applied Chemistry, School of Science, Northwestern Polytechnical University, Xi' an 710072 (China)

    2011-11-01

    Nickel plating graphite nanosheets (Ni/GNs) were prepared by electroless plating method using graphite nanosheets (GNs). Then a novel polymer magnetic composites based on acrylate pressure-sensitive adhesive (acrylate PSA) filled with Ni/GNs were fabricated by solution blend method. The Ni/GNs and acrylate PSA/Ni/GNs composites were characterized by scanning electron microscope (SEM)/energy dispersive spectroscopy (EDS), X-ray diffraction (XRD), vibrating sample magnetometer (VSM) and transmission electron microscope (TEM). All results showed that relatively uniform and compact Ni layer is successfully coated onto GNs under the given conditions, furthermore, Ni/GNs are homogeneously dispersed in acrylate PSA. The VSM results showed that the saturation magnetization of acrylate PSA/Ni/GNs composites increases with an increasing content of Ni/GNs while the coercivity decreases with an increasing content of Ni/GNs. When the content of GNs is 20 wt%, acrylate PSA/Ni/GNs composites exhibites good mechanical properties.

  1. Preparation and magnetic properties of polymer magnetic composites based on acrylate resin filled with nickel plating graphite nanosheets

    International Nuclear Information System (INIS)

    Zhang Yi; Qi Shuhua; Zhang Fan; Yang Yongqing; Duan Guochen

    2011-01-01

    Nickel plating graphite nanosheets (Ni/GNs) were prepared by electroless plating method using graphite nanosheets (GNs). Then a novel polymer magnetic composites based on acrylate pressure-sensitive adhesive (acrylate PSA) filled with Ni/GNs were fabricated by solution blend method. The Ni/GNs and acrylate PSA/Ni/GNs composites were characterized by scanning electron microscope (SEM)/energy dispersive spectroscopy (EDS), X-ray diffraction (XRD), vibrating sample magnetometer (VSM) and transmission electron microscope (TEM). All results showed that relatively uniform and compact Ni layer is successfully coated onto GNs under the given conditions, furthermore, Ni/GNs are homogeneously dispersed in acrylate PSA. The VSM results showed that the saturation magnetization of acrylate PSA/Ni/GNs composites increases with an increasing content of Ni/GNs while the coercivity decreases with an increasing content of Ni/GNs. When the content of GNs is 20 wt%, acrylate PSA/Ni/GNs composites exhibites good mechanical properties.

  2. Metal-functionalized single-walled graphitic carbon nitride nanotubes: a first-principles study on magnetic property

    Directory of Open Access Journals (Sweden)

    Shenoy Vivek

    2011-01-01

    Full Text Available Abstract The magnetic properties of metal-functionalized graphitic carbon nitride nanotubes were investigated based on first-principles calculations. The graphitic carbon nitride nanotube can be either ferromagnetic or antiferromagnetic by functionalizing with different metal atoms. The W- and Ti-functionalized nanotubes are ferromagnetic, which are attributed to carrier-mediated interactions because of the coupling between the spin-polarized d and p electrons and the formation of the impurity bands close to the band edges. However, Cr-, Mn-, Co-, and Ni-functionalized nanotubes are antiferromagnetic because of the anti-alignment of the magnetic moments between neighboring metal atoms. The functionalized nanotubes may be used in spintronics and hydrogen storage.

  3. 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

  4. Effect of formation temperature on properties of graphite/stannum composite for bipolar plate

    International Nuclear Information System (INIS)

    Selamat, Mohd Zulkefli; Yusuf, Muhammad Yusri Md; Wer, Tio Kok; Sahadan, Siti Norbaya; Malingam, Sivakumar Dhar; Mohamad, Noraiham

    2016-01-01

    Bipolar plates are key components in Proton Exchange Membrane (PEM) fuel cells. They carry current away from the cell and withstand the clamping force of the stack assembly. Therefore, PEM fuel cell bipolar plates must have high electrical conductivity and adequate mechanical strength, in addition to being light weight and low cost in terms of both applicable materials and production methods. In this research, the raw materials used to fabricate the high performance bipolar plate are Graphite (Gr), Stannum (Sn) and Polypropylene (PP). All materials used was in powder form and Gr and Sn act as fillers and the PP acts as binder. The ratio of fillers (Gr/Sn) and binder (PP) was fixed at 80:20. For the multi-conductive filler, small amount of Sn, which is 10 up to 20wt% (from the total weight of fillers 80%) have been added into Gr/Sn/PP composite. The fillers were mixed by using the ball mill machine. The second stage of mixing process between the mixer of fillers and binder is also carried out by using ball mill machine before the compaction process by the hot press machine. The effect of formation temperatures (160°C-170°C) on the properties of Gr/Sn/PP composite had been studied in detail, especially the electrical conductivity, bulk density, hardness and microstructure analysis of Gr/Sn/PP composite. The result shows that there are significant improvement in the electrical conductivity and bulk density, which are exceeding the US-DoE target with the maximum value of 265.35 S/cm and 1.682g/cm"3 respectively.

  5. Effect of formation temperature on properties of graphite/stannum composite for bipolar plate

    Energy Technology Data Exchange (ETDEWEB)

    Selamat, Mohd Zulkefli, E-mail: azulkeflis@utem.edu.my; Yusuf, Muhammad Yusri Md, E-mail: yusri.cheras@gmail.com; Wer, Tio Kok, E-mail: to91@hotmail.my; Sahadan, Siti Norbaya, E-mail: norbaya@utem.edu.my; Malingam, Sivakumar Dhar, E-mail: sivakumard@utem.edu.my; Mohamad, Noraiham, E-mail: noraiham@utem.edu.my [Centre of Advanced Research on Energy (CARe), Faculty of Mechanical Engineering, UniversitiTeknikal Malaysia Melaka, Hang Tuah Jaya, 76100 Durian Tunggal, Melaka (Malaysia)

    2016-03-29

    Bipolar plates are key components in Proton Exchange Membrane (PEM) fuel cells. They carry current away from the cell and withstand the clamping force of the stack assembly. Therefore, PEM fuel cell bipolar plates must have high electrical conductivity and adequate mechanical strength, in addition to being light weight and low cost in terms of both applicable materials and production methods. In this research, the raw materials used to fabricate the high performance bipolar plate are Graphite (Gr), Stannum (Sn) and Polypropylene (PP). All materials used was in powder form and Gr and Sn act as fillers and the PP acts as binder. The ratio of fillers (Gr/Sn) and binder (PP) was fixed at 80:20. For the multi-conductive filler, small amount of Sn, which is 10 up to 20wt% (from the total weight of fillers 80%) have been added into Gr/Sn/PP composite. The fillers were mixed by using the ball mill machine. The second stage of mixing process between the mixer of fillers and binder is also carried out by using ball mill machine before the compaction process by the hot press machine. The effect of formation temperatures (160°C-170°C) on the properties of Gr/Sn/PP composite had been studied in detail, especially the electrical conductivity, bulk density, hardness and microstructure analysis of Gr/Sn/PP composite. The result shows that there are significant improvement in the electrical conductivity and bulk density, which are exceeding the US-DoE target with the maximum value of 265.35 S/cm and 1.682g/cm{sup 3} respectively.

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

    International Nuclear Information System (INIS)

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

    1983-10-01

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

  7. A New Method to Measure Crack Extension in Nuclear Graphite Based on Digital Image Correlation

    Directory of Open Access Journals (Sweden)

    Shigang Lai

    2017-01-01

    Full Text Available Graphite components, used as moderators, reflectors, and core-support structures in a High-Temperature Gas-Cooled Reactor, play an important role in the safety of the reactor. Specifically, they provide channels for the fuel elements, control rods, and coolant flow. Fracture is the main failure mode for graphite, and breaching of the above channels by crack extension will seriously threaten the safety of a reactor. In this paper, a new method based on digital image correlation (DIC is introduced for measuring crack extension in brittle materials. Cross-correlation of the displacements measured by DIC with a step function was employed to identify the advancing crack tip in a graphite beam specimen under three-point bending. The load-crack extension curve, which is required for analyzing the R-curve and tension softening behaviors, was obtained for this material. Furthermore, a sensitivity analysis of the threshold value employed for the cross-correlation parameter in the crack identification process was conducted. Finally, the results were verified using the finite element method.

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

    International Nuclear Information System (INIS)

    Fachinger, Johannes

    2016-01-01

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

  9. 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.

  10. 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.

  11. Electrochemical properties of carbon nanocoils and hollow graphite fibers as anodes for rechargeable lithium ion batteries

    International Nuclear Information System (INIS)

    Wang, Liyong; Liu, Zhanjun; Guo, Quangui; Wang, Guizhen; Yang, Jinhua; Li, Peng; Wang, Xianglei; Liu, Lang

    2016-01-01

    Carbon nanocoils (CNCs) have been used as anode materials for preparation of lithium ion batteries. As pure carbon material without any chemical modification, the graphitized CNCs anode exhibited larger capacities with good Coulombic efficiency, a higher rate capability, and better reversibility than the hollow graphite fibers (HGFs) anode. The excellent performance of the CNCs was possibly ascribed to the special structure and the high degree of graphitization. As a result, the CNCs anode exhibited high reversible capacity of 385.5 mA h g"−"1 at 50 mA g"−"1, 104.7% reversible capacity retention after 105 cycles, and superior reversible capability of 177.4 mA h g"−"1 at 1 A g"−"1 after 100 cycles. This result indicated that CNCs could be an attractive choice as anode material for high-energy density and high-power lithium-ion batteries.

  12. 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.

  13. Pt nanoparticles embedded on reduced graphite oxide with excellent electrocatalytic properties

    Energy Technology Data Exchange (ETDEWEB)

    Saravanan, Gengan, E-mail: saravanan3che@gmail.com [Central University of Tamil Nadu, Department of Chemistry, Thiruvarur, 610101 (India); Mohan, Subramanian, E-mail: sanjnamohan@yahoo.com [EMFT Division, CSIR-Central Electrochemical Research Institute, Tamilnadu, Karaikudi 630 006 (India)

    2016-11-15

    Graphical abstract: RGO/Nano Pt: This study explore the electrocatalytic oxidation performance of reduced graphite oxide (RGO) anchored with nano Pt. This graphene composite reveal superior electrooxidation performance that is associated with the flexible RGO matrix and the uniform distribution of Pt particles, which enhances surface area, fast electron transfer, uniform particle size distribution; consequently, the RGO matrix provides more stability to Pt particles during electrooxidation process. Display Omitted - Highlights: • Greener electrochemical method applied to prepare well-dispersed Pt-rGO. • Pt-rGO large surface area excellent charge transfer better catalytic activity. • Low-cost highly efficient carbon-based electrodes for direct formic acid fuel cell. • rGO an excellent support to anchor Pt nanoparticles on its surface. • Pt-rGO distinctly enhanced current density towards formic acid electrooxidation. - Abstract: Economically viable electrochemical approach has been developed for the synthesis of Pt nanoparticles through electrodeposition technique on the surface of Reduced Graphite Oxide (RGO). Pt nanoparticles embedded Reduced Graphite Oxide on Glassy Carbon Electrode are employed (Pt-rGO/GCE) for electrooxidation of formic acid. Scanning Electron Microscopy (SEM) image and Transmission Electron Microscopy (TEM) image shows that reduced graphite oxide act as an excellent support to anchor the Pt nanoparticles. Cyclic voltammetry results confirmed that Pt-rGO/GCE enhanced current density as many folds than that of bare platinum electrode for electrooxidation of formic acid. X-ray diffraction (XRD) patterns for Pt-graphene composites illustrate that peaks at 69.15 and 23° for Pt (220) and graphene carbon (002) respectively. {sup 13}C NMR spectrum of the electrochemically reduced graphite oxide resonance contains only one peak at 133 ppm which retains graphitic sp{sup 2} carbon and does not contain any oxygenated carbon and the carbonyl

  14. 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

  15. Comparison of 3 MeV C+ ion-irradiation effects between the nuclear graphites made of pitch and petroleum cokes

    International Nuclear Information System (INIS)

    Chi, Se-Hwan; Kim, Gen-Chan

    2008-01-01

    Three million electron volt C + irradiation effects on the microstructure (crystallinity, crystal size), mechanical properties (hardness, Young's modulus) and oxidation of IG-110 (petroleum coke) and IG-430 (pitch coke) nuclear graphites were compared based on the materials characteristics (degree of graphitization (DOG), density, porosity, type of coke, Mrozowski cracks) of the grades and the ion-irradiation conditions. The specimens were irradiated up to ∼19 dpa at room temperature. Differences in the as-received microstructure were examined by Raman spectroscopy, X-ray diffraction (XRD), optical microscope (OM) and transmission electron microscope (TEM). The ion-induced changes in the microstructure, mechanical properties and oxidation characteristics were examined by the Raman spectroscopy, microhardness and Young's modulus measurements, and scanning electron microscope (SEM). Results of the as-received microstructure condition show that the DOG of the grades appeared the same at 0.837. The size of Mrozowski cracks appeared larger in the IG-110 of the higher open and total porosity than the IG-430. After an irradiation, the changes in the crystallinity and the crystallite size, both estimated by the Raman spectrum parameters, appeared large for the IG-430 and the IG-110, respectively. The hardness had increased after an irradiation, but, the hardness increasing behaviors were reversed at around 14 dpa. Thus, the IG-430 showed a higher increase before 14 dpa, but the IG-110 showed a higher increase after 14 dpa. No-clear differences in the increase of the Young's modulus were observed between the grades mainly due to a scattering in the measurements results. The IG-110 showed a higher oxidation rate than the IG-430 both before and after an irradiation. Besides the density and porosity, a possible contribution of the well-developed Mrozowski cracks in the IG-110 was noted for the observation. All the comparisons show that, even when the differences between the

  16. Comparison of 3 MeV C + ion-irradiation effects between the nuclear graphites made of pitch and petroleum cokes

    Science.gov (United States)

    Chi, Se-Hwan; Kim, Gen-Chan

    2008-10-01

    Three million electron volt C + irradiation effects on the microstructure (crystallinity, crystal size), mechanical properties (hardness, Young's modulus) and oxidation of IG-110 (petroleum coke) and IG-430 (pitch coke) nuclear graphites were compared based on the materials characteristics (degree of graphitization (DOG), density, porosity, type of coke, Mrozowski cracks) of the grades and the ion-irradiation conditions. The specimens were irradiated up to ˜19 dpa at room temperature. Differences in the as-received microstructure were examined by Raman spectroscopy, X-ray diffraction (XRD), optical microscope (OM) and transmission electron microscope (TEM). The ion-induced changes in the microstructure, mechanical properties and oxidation characteristics were examined by the Raman spectroscopy, microhardness and Young's modulus measurements, and scanning electron microscope (SEM). Results of the as-received microstructure condition show that the DOG of the grades appeared the same at 0.837. The size of Mrozowski cracks appeared larger in the IG-110 of the higher open and total porosity than the IG-430. After an irradiation, the changes in the crystallinity and the crystallite size, both estimated by the Raman spectrum parameters, appeared large for the IG-430 and the IG-110, respectively. The hardness had increased after an irradiation, but, the hardness increasing behaviors were reversed at around 14 dpa. Thus, the IG-430 showed a higher increase before 14 dpa, but the IG-110 showed a higher increase after 14 dpa. No-clear differences in the increase of the Young's modulus were observed between the grades mainly due to a scattering in the measurements results. The IG-110 showed a higher oxidation rate than the IG-430 both before and after an irradiation. Besides the density and porosity, a possible contribution of the well-developed Mrozowski cracks in the IG-110 was noted for the observation. All the comparisons show that, even when the differences between the

  17. 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

  18. 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)

  19. Comparison of fracture toughness (K{sub IC}) and strain energy release rate (G) of selected nuclear graphites

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Se-Hwan, E-mail: shchi@kaeri.re.kr

    2016-08-01

    The fracture behaviors of six nuclear graphite grades for a high-temperature gas-cooled reactor (HTGR), which differed in coke particle size and forming method, were characterized based on the ASTM standard graphite fracture toughness test method (ASTM D 7779-11) at room temperature. The G appeared to show good correlation with the fracture surface roughness and the G-Δa curves appeared to describe the fracture process well from crack initiation to failure. Comparison of the local (K{sub IC}) and gross (G{sub IC}, G-Δa) fracture parameters showed that the resistance to crack initiation and propagation was higher in the extruded or vibration molded medium particle size grades (PCEA, NBG-17, NBG-18: EVM group) than in the iso-molded fine particle size grades (IG-110, IG-430, NBG-25: IMF group). The ASTM may need to provide a guideline for G-Δa curve analysis. The K{sub IC} appeared to increase with specimen thickness (size).

  20. Isospin dependent properties of asymmetric nuclear matter

    OpenAIRE

    Chowdhury, P. Roy; Basu, D. N.; Samanta, C.

    2009-01-01

    The density dependence of nuclear symmetry energy is determined from a systematic study of the isospin dependent bulk properties of asymmetric nuclear matter using the isoscalar and the isovector components of density dependent M3Y interaction. The incompressibility $K_\\infty$ for the symmetric nuclear matter, the isospin dependent part $K_{asy}$ of the isobaric incompressibility and the slope $L$ are all in excellent agreement with the constraints recently extracted from measured isotopic de...

  1. The effect of iron catalyzed graphitization on the textural properties of carbonized cellulose : Magnetically separable graphitic carbon bodies for catalysis and remediation

    NARCIS (Netherlands)

    Hoekstra, Jacco; Beale, Andrew M.; Soulimani, Fouad; Versluijs-Helder, Marjan; Van De Kleut, Dirk; Koelewijn, Jacobus M.; Geus, John W.; Jenneskens, Leonardus W.

    2016-01-01

    Whereas pyrolysis of pristine microcrystalline cellulose spheres yields nonporous amorphous carbon bodies, pyrolysis of microcrystalline cellulose spheres loaded with iron salts leads to the formation of magnetically separable mesoporous graphitic carbon bodies. The microcrystalline cellulose

  2. Effect of extensional cyclic strain on the mechanical and physico-mechanical properties of PVC-NBR/graphite composites

    Directory of Open Access Journals (Sweden)

    2008-12-01

    Full Text Available The variation of electrical resistivity as will as the mechanical properties of PVC (polyvinylchloride-NBR (acrylonitrile butadiene rubber based conductive composites filled with different concentrations of graphite were studied. These samples were studied as function of the constant deformation fatigue test. When the specimen was subjected to a large number of rapidly repeating strain cycles, and different strain amplitudes, the conductivity, σ(T, shows an initial rapid fall followed by dynamic equilibrium. Increasing the number of cycles and strain amplitudes, the conductivity remains almost constant over the temperature range 30–140°C. The equilibrium state between destruction and reconstruction of graphite particles has been detected for all strains of certain values of strain cycles (1000, 2000, 3000, and 4000 cycles for 30% strain amplitude. A preliminary study was done to optimize the possibility to use Conductive Polymer Composites (CPC as a strain sensor and to evaluate its performance by an intrinsic physico-mechanical modification measurement. The electromechanical characterization was performed to demonstrate the adaptability and the correct functioning of the sensor as a strain gauge on the fabric. The coefficient of strain sensitivity (K was measured for 50 phr graphite/PVCNBR vulcanized at 3000 number of strain cycles and 30% strain amplitude. There was a broad maximum of K, with a peak value of 82, which was much higher, compared to conventional wire resistors. A slight hysteresis was observed at unloading due to plasticity of the matrix. A good correlation exists between mechanical and electrical response to the strain sensitivity. Mechanical reinforcement was in accordance with the Quemada equation [1] and Guth model [2] attested to good particle-matrix adhesion. It was found that the viscous component of deformation gradually disappeared and the hardening occurred with increasing strain cycles. The modulus, fracture

  3. High-temperature irradiation effects on mechnical properties of HTGR graphites

    International Nuclear Information System (INIS)

    Oku, Tatsuo; Eto, Motokuni; Fujisaki, Katsuo

    1978-04-01

    The irradiation effects on stress-strain relation, Young's modulus, tensile strength, bending strength and compressive strength of HTGR graphites were studied in irradiation temperature ranges of 200 - 300 0 C and 800 - 1400 0 C and in neutron fluences up to 7.4 x 10 20 n/cm 2 and 3 x 10 21 n/cm 2 (> 0.18 MeV). Fracture criteria and strain energy to fracture of the unirradiated and the irradiated graphites were also examined. (1) Neutron fluence dependences are similar in Young's modulus, tensile strength and bending strength. (2) The change of compressive strength and of tensile and bending strengths with neutron fluence differ; the former varies with graphite kind. (3) At lower irradiation temperatures the bending fracture strain energy decreases with increasing neutron fluence and at higher irradiation temperatures it increases. (4) The fracture criteria of graphites deviates from the constant strain energy theory (α = 0.5) and the constant strain theory (α = 1), shifting from α asymptotically equals 0.5 to α asymptotically equals 1 with increasing irradiation temperature. (auth.)

  4. Effect of graphite nanoplatelets on the structure and properties of PA6-elastomer nanocomposites

    Czech Academy of Sciences Publication Activity Database

    Thanh, Tran Duy; Kaprálková, Ludmila; Hromádková, Jiřina; Kelnar, Ivan

    2014-01-01

    Roč. 50, January (2014), s. 39-45 ISSN 0014-3057 R&D Projects: GA ČR(CZ) GA13-15255S Institutional support: RVO:61389013 Keywords : graphite nanoplatelets * polyamide 6 * elastomer Subject RIV: JI - Composite Materials Impact factor: 3.005, year: 2014

  5. 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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  7. Average Nuclear properties based on statistical model

    International Nuclear Information System (INIS)

    El-Jaick, L.J.

    1974-01-01

    The rough properties of nuclei were investigated by statistical model, in systems with the same and different number of protons and neutrons, separately, considering the Coulomb energy in the last system. Some average nuclear properties were calculated based on the energy density of nuclear matter, from Weizsscker-Beth mass semiempiric formulae, generalized for compressible nuclei. In the study of a s surface energy coefficient, the great influence exercised by Coulomb energy and nuclear compressibility was verified. For a good adjust of beta stability lines and mass excess, the surface symmetry energy were established. (M.C.K.) [pt

  8. Nuclear properties for astrophysical applications

    Energy Technology Data Exchange (ETDEWEB)

    Moeller, P.; Nix, J.R. [Los Alamos National Lab., NM (United States); Kratz, K.L. [Mainz Univ. (Germany). Inst. fuer Kernchemie

    1994-09-23

    We tabulate the ground-state odd-proton and odd-neutron spins, proton and neutron pairing gaps, binding energies, neuton separation energies, quantities related to {beta}-delayed one, two and three neutron emission probabilities, {beta}-decay Q values and half-lives with respect to Gamow-Teller decay, proton separation energies, and {alpha}-decay Q values and half-lives. The starting point of the calculations is a calculation of nuclear ground-states and (information based on the finite-range droplet model and the folded-Yukawa single-particle model published in a previous issue of ATOMIC DATA AND NUCLEAR DATA TABLES. The {beta}-delayed neutron-emission probabilities and Gamow-Teller {beta}-decay rates are obtained from a QRPA model that uses single-particle levels and wave-functions at the calculated nuclear ground-state shape as the starting point.

  9. Nitrogen-doped graphene: effect of graphite oxide precursors and nitrogen content on the electrochemical sensing properties.

    Science.gov (United States)

    Megawati, Monica; Chua, Chun Kiang; Sofer, Zdenek; Klímová, Kateřina; Pumera, Martin

    2017-06-21

    Graphene, produced via chemical methods, has been widely applied for electrochemical sensing due to its structural and electrochemical properties as well as its ease of production in large quantity. While nitrogen-doped graphenes are widely studied materials, the literature showing an effect of graphene oxide preparation methods on nitrogen quantity and chemical states as well as on defects and, in turn, on electrochemical sensing is non-existent. In this study, the properties of nitrogen-doped graphene materials, prepared via hydrothermal synthesis using graphite oxide produced by various classical methods using permanganate or chlorate oxidants Staudenmaier, Hummers, Hofmann and Brodie oxidation methods, were studied; the resulting nitrogen-doped graphene oxides were labeled as ST-GO, HU-GO, HO-GO and BR-GO, respectively. The electrochemical oxidation of biomolecules, such as ascorbic acid, uric acid, dopamine, nicotinamide adenine nucleotide and DNA free bases, was carried out using cyclic voltammetry and differential pulse voltammetry techniques. The nitrogen content in doped graphene oxides increased in the order ST-GO graphene followed this trend, as shown in the cyclic voltammograms. This is a very important finding that provides insight into the electrocatalytic effect of N-doped graphene. The nitrogen-doped graphene materials exhibited improved sensitivity over bare glassy carbon for ascorbic acid, uric acid and dopamine detection. These studies will enhance our understanding of the effects of graphite oxide precursors on the electrochemical sensing properties of nitrogen-doped graphene materials.

  10. Experimental measurement of nuclear heating in a graphite-cantered assembly in deuterium-tritium neutron environment for the validation of data and calculation

    International Nuclear Information System (INIS)

    Kumar, A.; Youssef, M.; Abdou, M.A.

    1998-01-01

    Within the framework of the ITER Task T-218 entitled 'Shielding Blanket Neutronics Experiments', nuclear heating measurements were conducted jointly by the USA and Japan using a micro calorimetric technique in a graphite-cantered assembly. An accelerator-based D-T neutron source at JAERI was used to provide a mixed neutron and photon field. The first measurements related to direct micro calorimetric measurements in individual graphite probes along the axis. In the second set, the first graphite probe was replaced, one by one, by single probes of beryllium, aluminum, silicon, silicon carbide, titanium, vanadium, chromium, iron, stainless steel 316, nickel, copper, zirconium, niobium, molybdenum, tungsten. Analysis of the measurements has been carried out using Monte Carlo code MCNP with FENDL-1, ENDF/B-VI and MCPLIB nuclear data libraries. A comparison of calculations (C) and experiments (E) shows a C/E ratio lying in a C/E band extending from 0.9 to 1.2 for beryllium, graphite, copper, chromium, iron, nickel, 316 stainless steel, titanium, vanadium, molybdenum, niobium and tungsten. However, larger deviations from unity are seen for C/E values for silicon, zirconium, and aluminum. Though FENDL-1 and ENDF/B-VI libraries provide very close nuclear heating rates for most of the probe materials, significant divergences are seen for silicon, silicon carbide, aluminum, titanium, zirconium, niobium, and molybdenum. The divergences are traceable to differences in neutron kerma factors as well as gamma production cross-sections of these materials. (orig.)

  11. Electric and magnetic properties of the stage-2 FeBr{sub 2} graphite intercalation compound

    Energy Technology Data Exchange (ETDEWEB)

    Dube, P A; Barati, M; Ummat, P K; Luke, G; Datars, W R [Department of Physics and Astronomy, McMaster University, Hamilton, ON (Canada)

    2003-01-22

    The stage-2 FeBr{sub 2} graphite intercalation compound (GIC) was prepared by reacting FeBr{sub 2} powder and highly oriented pyrolytic graphite in a bromine atmosphere at 500 deg. C for 40 weeks. The dc magnetization, ac susceptibility, specific heat, resistivity and Hall effect were measured. The GIC is paramagnetic at temperatures above 14.5 K. There is short-range ordering at 14.5 K and longer-range magnetic ordering at 8.5 K. There is a spin glass phase below 3.2 K in which the ac susceptibility is frequency dependent. The in-plane and c-axis resistivities result from in-plane and out-of-plane electron-phonon scattering. The Hall coefficient is independent of temperature between 4.2 and 300 K and is explained by the single-carrier model.

  12. Electric and magnetic properties of the stage-2 FeBr2 graphite intercalation compound

    International Nuclear Information System (INIS)

    Dube, P A; Barati, M; Ummat, P K; Luke, G; Datars, W R

    2003-01-01

    The stage-2 FeBr 2 graphite intercalation compound (GIC) was prepared by reacting FeBr 2 powder and highly oriented pyrolytic graphite in a bromine atmosphere at 500 deg. C for 40 weeks. The dc magnetization, ac susceptibility, specific heat, resistivity and Hall effect were measured. The GIC is paramagnetic at temperatures above 14.5 K. There is short-range ordering at 14.5 K and longer-range magnetic ordering at 8.5 K. There is a spin glass phase below 3.2 K in which the ac susceptibility is frequency dependent. The in-plane and c-axis resistivities result from in-plane and out-of-plane electron-phonon scattering. The Hall coefficient is independent of temperature between 4.2 and 300 K and is explained by the single-carrier model

  13. Study of Coating Geometries and Photoluminescence Properties of Metal Nanoparticles/Graphite Composites

    Directory of Open Access Journals (Sweden)

    Pasquale Barone

    2014-01-01

    Full Text Available In this work we present the results of a study of growth and characterization of metal nanoparticles (Ag, Au, and Co/carbon surfaces. The nanoparticles grew by laser ablation technique and their dimensions were controlled by light scattering study and AFM microscopy before their insertion on graphite surface. Nanoparticles appear randomly disposed on carbon surfaces aggregating to form big particles only in the case of silver. The different behavior of metal nanoparticles on carbon surface was explained in terms of different metal wetting of surface, in agreement with previous theoretical results of He et al. Chemical information, obtained by X-ray photoelectron spectroscopy, indicated that the doping process is a simple physisorption while the interfacial interaction between particles and carbon layers causes local defects in graphite structure and the appearance of a strong photoluminescence signal for all composites. Moreover, the visible optical absorption decreases about 10% indicating the progressive metallization of carbon surface.

  14. Innovative approaches to the Management of Irradiated Nuclear Graphite Wastes: Addressing the Challenges through International Collaboration with Project 'GRAPA'

    International Nuclear Information System (INIS)

    Wickham, A.J.; Ojovan, M.; O'Sullivan, P.; )

    2017-01-01

    There exists more than 250.000 tonnes of irradiated (and therefore radioactive) nuclear graphite (i-graphite) in the world, primarily as a result of the development of graphite-moderated power-reactor systems, initially for defence and subsequently for commercial purposes. Only a very small number of such plants have been dismantled and, for most cases, the final destiny of the irradiated graphite remains unresolved. Future high-temperature reactor programmes, such as the Chinese HTR-PM development, will produce more graphite and carbonaceous wastes from both structural components and the fuel pebbles (which are approximately 96% carbonaceous), the latter producing a continuous stream of so-called 'operational waste'. The problem of dismantling irradiated graphite reactor stacks, possibly distorted through neutron damage and in some cases degraded further by radiation-chemical attack by gaseous coolants, and then finding the appropriate treatments and final destiny of the material, has exercised both the European Union and the International Atomic Energy Agency for more than 25 years, seeking to address the different issues and available disposal solutions in different IAEA Member States. An IAEA collaborative research programme on treatment options has recently been completed, and an active group of international specialists in this area has now been established as part of the IAEA International Decommissioning Network under the envelope of Project 'GRAPA' (Irradiated Graphite Processing Approaches), which includes representatives from Belgium, China, France, Germany, India, Italy, Lithuania, Rep. of Korea, Romania, Spain, Switzerland, Ukraine and the Russian Federation with direct responsibilities for various parts of the decommissioning and graphite-disposal process in a variety of reactor designs. Interest has also been expressed by colleagues from Sweden and Japan. Work is in progress on a number of topic areas where weaknesses in the

  15. Studies of mechanical properties and irradiation damage nucleation of HTGR graphites. Final report

    International Nuclear Information System (INIS)

    Thrower, P.A.

    1981-05-01

    Since the submission of the last report (COO-2712-6) work has concentrated on the examination of the effects of oxidation on the compressive strengths of graphites doped with iron, vanadium and calcium. The purpose of the investigation was to determine the relative effects of the impurities on the rates of oxidation in air, CO 2 and H 2 O and the resultant reduction in compressive strength

  16. Heat transfer in the core graphite structures of RBMK nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Knoglinger, E., E-mail: ernst.knoglinger@a1.net [Am Winklerwald 15, A 4020 Linz (Austria); Wölfl, H., E-mail: herbert.woelfl@tele2.at [Berg, Im Weideland 19, A 4060 Linz (Austria); Kaliatka, A., E-mail: algirdas.kaliatka@lei.lt [Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos 3, LT-44403 Kaunas (Lithuania)

    2015-11-15

    Highlights: • Proposed solution of heat transfer model from a hollow cylinder to a fluid through narrow duct. • Thermal conductance of annular gaps, filled by two component gas was discussed. • Xenon transient preceding the Chernobyl Accident was analyzed. • Reactivity balance during power manoeuvres and potenrial causes of the accident were discussed. - Abstract: Conductive and combined radiative/conductive gap conductance models are presented and discussed in great detail. The heat resistance concept and an exact solution to the one dimensional heat conduction equation for a 3-region composite hollow cylinder are used to calculate gap conductance in function of gap gas composition and fuel burn up. The study includes the back calculation of a reactor experiment performed at the Ignalina NPP Unit-1 which provides some insight in the function of the RBMK nitrogen supply and regulating device and an investigation of the role the graphite temperature played during the power manoeuvres preceding the Chernobyl Accident.

  17. 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.

  18. Simultaneously improving the mechanical and electrical properties of poly(vinyl alcohol) composites by high-quality graphitic nanoribbons.

    Science.gov (United States)

    Yang, Ming; Weng, Lin; Zhu, Hanxing; Zhang, Fan; Fan, Tongxiang; Zhang, Di

    2017-12-07

    Although carbon nanotubes (CNTs) have shown great potential for enhancing the performance of polymer matrices, their reinforcement role still needs to be further improved. Here we implement a structural modification of multi-walled CNTs (MWCNTs) to fully utilize their fascinating mechanical and electrical properties via longitudinal splitting of MWCNTs into graphitic nanoribbons (GNRs). This nanofiller design strategy is advantageous for surface functionalization, strong interface adhesion as well as boosting the interfacial contact area without losing the intrinsic graphitic structure. The obtained GNRs have planar geometry, quasi-1D structure and high-quality crystallinity, which outperforms their tubular counterparts, delivering a superior load-bearing efficiency and conductive network for realizing a synchronous improvement of the mechanical and electrical properties of a PVA-based composite. Compared to PVA/CNTs, the tensile strength, Young's modulus and electrical conductivity of the PVA/GNR composite at a filling concentration of 3.6 vol.% approach 119.1 MPa, 5.3 GPa and 2.4 × 10 -4 S m -1 , with increases of 17%, 32.5% and 5.9 folds, respectively. The correlated mechanics is further rationalized by finite element analysis, the generalized shear-lag theory and the fracture mechanisms.

  19. Temperature dependence of nuclear surface properties

    International Nuclear Information System (INIS)

    Campi, X.; Stringari, S.

    1982-01-01

    Thermal properties of nuclear surface are investigated in a semi-infinite medium. Explicit analytical expression are given for the temperature dependence of surface thickness, surface energy and surface free energy. In this model the temperature effects depend critically on the nuclear incompressibility and on the shape of the effective mass at the surface. To illustrate the relevance of these effects we made an estimate of the temperature dependence of the fission barrier height. (orig.)

  20. Ignition properties of nuclear grade activated carbons

    International Nuclear Information System (INIS)

    Freeman, W.P.; Hunt, J.R.; Kovach, J.L.

    1983-01-01

    The ignition property of new activated carbons used in air cleaning systems of nuclear facilities has been evaluated in the past, however very little information has been generated on the behavior of aged, weathered carbons which have been exposed to normal nuclear facility environment. Additionally the standard procedure for evaluation of ignition temperature of carbon is performed under very different conditions than those used in the design of nuclear air cleaning systems. Data were generated evaluating the ageing of activated carbons and comparing their CH 3 131 I removal histories to their ignition temperatures. A series of tests were performed on samples from one nuclear power reactor versus use time, a second series evaluated samples from several plants showing the variability of atmospheric effects. The ignition temperatures were evaluated simulating the conditions existing in nuclear air cleaning systems, such as velocity, bed depth, etc., to eliminate potential confusion resulting from artifically set current standard conditions

  1. Influence of Fenton's reagent treatment on electrochemical properties of graphite felt for all vanadium redox flow battery

    International Nuclear Information System (INIS)

    Gao, Chao; Wang, NanFang; Peng, Sui; Liu, SuQin; Lei, Ying; Liang, XinXing; Zeng, ShanShan; Zi, HuiFang

    2013-01-01

    Highlights: ► Highly hydroxyl-functionalized graphite felt has been obtained through Fenton's reagent treatment. ► Fenton's reagent treatment involves only one step, works under ambient conditions and will never produce any toxic gas. ► The treated graphite felt exhibits superior electrochemical performance in comparison to the untreated one. -- Abstract: An environmental, economic and highly effective method for carbon fiber hydroxylated-functionalization based on Fenton's reagent treatment is used to improve the electrochemical activity of graphite felt (GF) as the positive electrode in all vanadium redox flow battery (VRFB). The effect of H 2 O 2 content in Fenton's reagent on the structure and electrochemical properties of GF is investigated. The scanning electron microscope (SEM) indicates that the surface of the treated GF is etched increasingly with the content of H 2 O 2 . The Fourier transformation infrared (FTIR) spectroscopy shows that the peak intensity of hydroxyl groups on the treated felt is increased with the H 2 O 2 concentration, which is further verified by X-ray photoelectron spectroscopy (XPS). Cyclic voltammetry (CV) and electrochemical impedance spectroscopy (EIS) show that the treated sample exhibits a higher electrochemical activity. A VRFB with the treated GF as electrodes exhibits higher coulombic, voltage and energy efficiency (98.8%, 75.1% and 74.2%) than that with the untreated GF (93.9%, 72.1% and 67.7%) at 60 mA cm −2 , and this method is even superior when compared with the reported methods

  2. Nuclear medium effects on the nucleon properties

    International Nuclear Information System (INIS)

    Gerard, A.

    1986-06-01

    The idea that the nuclear medium may modify the properties of the nucleon is not really a new one. But until these last few years the intrinsic properties of the nucleon such as its mass, baryonic or electromagnetic size, magnetic moments, were hardly suspected to be substantially modified by the nuclear environment. The question raised essentially from two categories of experimental investigations, in quite different kinematical regimes and it remains at the present time a controversal subject. In this paper I try to review the present status of experimental investigations and some of the recent theoretical developments relevant to this topic

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  4. 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

  5. The mechanical properties and morphology of a graphite oxide nanoplatelet/polyurethane composite

    International Nuclear Information System (INIS)

    Cai Dongyu; Yusoh, Kamal; Song Mo

    2009-01-01

    Significant reinforcement of polyurethane (PU) using graphite oxide nanoplatelets (GONPs) is reported. Morphologic study shows that, due to the formation of chemical bonding, there is a strong interaction between the GONPs and the hard segment of the PU, which allows effective load transfer. The GONPs can prevent the formation of crystalline hard segments due to their two-dimensional structure. With the incorporation of 4.4 wt% of GONPs, the Young's modulus and hardness of the PU are significantly increased by ∼900% and ∼327%, respectively. The resultant high resistance to scratching indicates promise for application of these composite materials in surface coating.

  6. AGC 2 Irradiated Material Properties Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Rohrbaugh, David Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-05-01

    The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

  7. Thorium dioxide: properties and nuclear applications

    International Nuclear Information System (INIS)

    Belle, J.; Berman, R.M.

    1984-01-01

    This is the sixth book on reactor materials published under sponsorship of the Naval Reactors Office of the United States Department of Energy, formerly the United States Atomic Energy Commission. This book presents a comprehensive compilation of the most significant properties of thorium dioxide, much like the book Uranium Dioxide: Properties and Nuclear Applications presented information on the fuel material used in the Shippingport Pressurized Water Reactor core

  8. Thorium dioxide: properties and nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Belle, J.; Berman, R.M. (eds.)

    1984-01-01

    This is the sixth book on reactor materials published under sponsorship of the Naval Reactors Office of the United States Department of Energy, formerly the United States Atomic Energy Commission. This book presents a comprehensive compilation of the most significant properties of thorium dioxide, much like the book Uranium Dioxide: Properties and Nuclear Applications presented information on the fuel material used in the Shippingport Pressurized Water Reactor core.

  9. Measurements of Chlorine by XPS and SIMS in Nuclear-Grade Graphite. Effect of Treatment by UV, Heat and Moisture

    International Nuclear Information System (INIS)

    Grossiord, C.; Amalric, J.; Rahmani, L.

    2016-01-01

    Pieces of historical nuclear-grade graphite from the same block were submitted to a treatment involving soft or hard UV and heating in moist air. XPS investigations before treatment have shown that – at the geometric surface at least – chlorine binds more commonly by covalent (30 to 100% depending on spots) than ionic bonds. After treatment, bond types are more balanced, though hardly more uniform. Shifts of the ionic peak indicating an oxidised bond were not found. Surface chlorine was measured up to 0.04% atomic. Treatment by hard UV and heating in moist air for 1 ½ hours up to 122°C led to a decrease of surface chlorine by a factor of 2 as measured by ToF and XPS, helped by an etching technique that ensured the accurate positioning of the spots under analysis. Treatment for 8 hours up to 151°C resulted in a decrease of the ToF Cl/ 13 C signal for surface chlorine from 0.73 to 0.20 and of the atomic fraction from 0.02% to 0% as measured by XPS. SIMS investigations reveal that in the first 6 μm from the geometric surface, chlorine, hydrogen and oxygen have a decreasing profile suggestive of diffusion from the surface. In the vicinity to the surface the amounts of all three elements are much greater than in the bulk, possibly due to the availability of dangling bonds. It is speculated that when these are not saturated, back diffusion from the bulk under treatment may increase the quantity of surface chlorine. Treatment in moist air by hard UV for to 1 ½ hours has no marked effect on chlorine in the bulk, although it alters the quantities of hydrogen and oxygen. Treatment for 8 hours decreases chlorine by a factor of about 5 throughout the analysed depth of 6 μm and possibly farther to 15 μm, suggesting its diffusivity to be at least an order greater than 3.10 -16m 2 /s and possibly than 2.10 -15 m²/s, although a range of values is expected owing to the diverse tortuosity and constrictivity of pores. Ratios of hydrogen to oxygen and of chlorine to

  10. 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.

  11. 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

  12. Enhanced photovoltaic properties in graphitic carbon nanospheres networked TiO{sub 2} nanocomposite based dye sensitized solar cell

    Energy Technology Data Exchange (ETDEWEB)

    Agarwal, Radhe [Department of Physics and Institute for Functional Nanomaterials, University of Puerto Rico, San Juan, PR 00931 (United States); Sahoo, Satyaprakash, E-mail: satya504@gmail.com [Department of Physics and Institute for Functional Nanomaterials, University of Puerto Rico, San Juan, PR 00931 (United States); Chitturi, Venkateswara Rao [Department of Physics and Institute for Functional Nanomaterials, University of Puerto Rico, San Juan, PR 00931 (United States); Williams, Joseph D. [Department of Biomedical and Chemical Engineering, Syracuse University, L.C. Smith College of Engineering and Computer Science, Syracuse, NY (United States); Resto, Oscar [Department of Physics and Institute for Functional Nanomaterials, University of Puerto Rico, San Juan, PR 00931 (United States); Katiyar, Ram S., E-mail: rkatiyar@hpcf.uprrp.edu [Department of Physics and Institute for Functional Nanomaterials, University of Puerto Rico, San Juan, PR 00931 (United States)

    2015-08-25

    Highlights: • Nano size graphitic carbon nanospheres were prepared from MWCNTs. • TiO{sub 2}/GCNS composite was used as the photoanode in dye-sensitized solar cell. • An improved photovoltaic performance with GCNS–TiO{sub 2} composite was noticed. - Abstract: In this work, we report a novel carbon based TiO{sub 2} nanocomposite electron injection layer (photoanode) toward the improved performance of DSSCs. Graphitic carbon nanospheres (GCNSs) were synthesized by a unique acidic treatment of multi-wall carbon nanotubes. GCNS–TiO{sub 2} nanocomposites with different concentrations of GCNSs (ranging from 5 to 20 μL) were prepared to use as photoanodes in DSSCs. Structural and morphological properties of GCNS–TiO{sub 2} nanocomposites were analyzed by Raman spectroscopy and ultra-high resolution transmission electron microscopy techniques, respectively. A systematic increment in the short circuit current density (J{sub SC}) and open circuit voltage (V{sub OC}) of DSSC was observed by increasing GCNS concentration up to an optimal value, possibly due to the combined effect of slight rise in quasi-Fermi level and higher carrier transport rate in the resultant composite. Thus, a significant enhancement of ∼47% in the efficiency of DSSC containing GCNS–TiO{sub 2} photoanode was observed as compare to DSSC with pure TiO{sub 2} photoanode.

  13. Electrocatalytic properties of N-doped graphite felt in electro-Fenton process and degradation mechanism of levofloxacin.

    Science.gov (United States)

    Liu, Xiaocheng; Yang, Danxing; Zhou, Yaoyu; Zhang, Jiachao; Luo, Lin; Meng, Sijun; Chen, Song; Tan, Mengjiao; Li, Zhicheng; Tang, Lin

    2017-09-01

    The degradation of antibiotic levofloxacin was investigated by dimensionally stable anode as well as modified cathode using low-cost chemical reagents of hydrazine hydrate and ethanol for electro-Fenton in an undivided cell at pH 3.0 under room temperature. Comparison of unmodified and modified cathode was performed. The apparent rate constant of levofloxacin decay was found to be 0.2883 min -1 for graphite felt-10 with the best performance at 200 mA, which is lower than graphite felt at 400 mA. The optimum modified cathode showed a significant improvement of complete mineralization of levofloxacin, reaching a 92% TOC removal at 200 mA for 480 min higher than unmodified one at twice the current. Surface physicochemical properties and morphology were investigated by scanning electron microscope, contact angle and X-ray photoelectron spectroscopy. The electrochemical characterization of hydrogen evolution reaction was adopted to clarify a possible pathway for the higher mineralization of levofloxacin, indicating a potential pilot-scale study to the pollution with the similar structure. Copyright © 2017 Elsevier Ltd. All rights reserved.

  14. Liability for on-site nuclear property damage

    International Nuclear Information System (INIS)

    Neems, H.J.

    2000-01-01

    Typically, liability for on-site property addressed in contracts between operator and its suppliers. Nuclear power plant operators ordinarily protect themselves against risk of nuclear damage to on-site property by insurance. Nuclear liability laws do not specifically address liability for nuclear damage to on-site property. Nuclear plant owners should address risk of damage to on-site property when developing risk management program

  15. Isospin dependent properties of asymmetric nuclear matter

    Science.gov (United States)

    Chowdhury, P. Roy; Basu, D. N.; Samanta, C.

    2009-07-01

    The density dependence of nuclear symmetry energy is determined from a systematic study of the isospin dependent bulk properties of asymmetric nuclear matter using the isoscalar and isovector components of the density dependent M3Y interaction. The incompressibility K∞ for the symmetric nuclear matter, the isospin dependent part Kasy of the isobaric incompressibility, and the slope L are all in excellent agreement with the constraints recently extracted from measured isotopic dependence of the giant monopole resonances in even-A Sn isotopes, from the neutron skin thickness of nuclei, and from analyses of experimental data on isospin diffusion and isotopic scaling in intermediate energy heavy-ion collisions. This work provides a fundamental basis for the understanding of nuclear matter under extreme conditions and validates the important empirical constraints obtained from recent experimental data.

  16. Characterization of un-irradiated and irradiated reactor graphite; Karakterizacija neozracenog i ozracenog reaktorskog grafita

    Energy Technology Data Exchange (ETDEWEB)

    Marinkovic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    This report contains three parts: characterization of Yugoslav nuclear graphite development of methods and obtained results, characterization of un-irradiated and irradiated domestic nuclear graphite; calculation of electrical conductivity changes due to vacancies in the graphite crystal lattice.

  17. Temperature and radiolytic corrosion effects on the chlorine behaviour in nuclear graphite: consequences for the disposable of irradiated graphite from UNGG reactors

    International Nuclear Information System (INIS)

    Vaudey, C.E.

    2010-10-01

    This work concerns the dismantling of the UNGG reactor which have produced around 23 000 t of graphite wastes that ave to be disposed of according to the French law of June 206. These wastes contain two long-lived radionuclides ( 14 C and 36 Cl) which are the main long term dose contributors. In order to get information about their inventory and their long term behaviour in case of water ingress into the repository, it is necessary to determine their location and speciation in the irradiated graphite after the reactor shutdown. This work concerns the study of 36 Cl. The main objective is to reproduce its behaviour during reactor operation. For that purpose, we have studied the effects of temperature and radiolytic corrosion independently. Our results show a rapid release of around 20% 36 Cl during the first hours of reactor operation whereas a much slower release occurs afterwards. We have put in evidence two types of chlorine corresponding to two different chemical forms (of different thermal stabilities) or to two locations (of different accessibilities). We have also shown that the radiolytic corrosion seems to enhance chlorine release, whatever the irradiation dose. Moreover, the major chemical form of chlorine is inorganic. (author)

  18. Effects of specimen size on the flexural strength and Weibull modulus of nuclear graphite IG-110, NBG-18, and PCEA

    International Nuclear Information System (INIS)

    Chi, Se-Hwan

    2015-01-01

    Changes in flexural strength and Weibull modulus due to specimen size were investigated for three nuclear graphite grades, IG-110, NBG-18, and PCEA, using four-point-1/3 point (4-1/3) loading with specimens of three different sizes: 3.18 (Thickness) × 6.35 (Width) × 50.8 (Length), 6.50 (T) × 12.0 (W) × 52.0 (L), 18.0 (T) × 16.0 (W) × 64 (L) (mm) (total: 210 specimens). Results showed some specimen size effects were grade dependent: While NBG-18 (a) showed rather significant specimen size effects (37% difference between the 3 T and 18 T), the differences in IG-110 and PCEA were 7.6–15%. The maximum differences in flexural strength due to specimen size were larger in the PCEA and NBG-18 having larger sized coke particles (medium grain size: >300 μm) than the IG-110 with super fine coke particle size (25 μm). The Weibull modulus showed a data population dependency, in that it decreased with increasing numbers of data used for modulus determination. A good correlation between the fracture surface roughness and the flexural strength was confirmed

  19. Study on the property of low friction complex graphite-like coating containing tantalum

    Science.gov (United States)

    Wang, Zuoping; Feng, Lajun; Shen, Wenning

    2018-03-01

    In order to enhance equipment lifetime under low oil or even dry conditions, tantalum was introduced into the graphite-like coating (GLC) by sputtering mosaic targets. The results showed that the introduction of Ta obviously reduced the friction coefficient and hardness of the GLC, while improved the wearability. When the atomic percentage of Ta was larger than 3%, the steady friction coefficient was lower than 0.01, suggesting the coating exhibited super lubricity. When the content of Ta was about 5.0%, the average friction coefficient was 0.02 by a sliding friction test under load of 20 N in unlubricated condition. Its average friction coefficient reduced by 75%, compared with that of control GLC (0.0825).

  20. 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.

  1. Safety problems of nuclear power plants with channel-type graphite boiling water reactors

    International Nuclear Information System (INIS)

    Emel'yanov, I.Ya.; Vasilevskij, V.P.; Volkov, V.P.; Gavrilov, P.A.; Kramerov, A.Ya.; Kuznetsov, S.P.; Kunegin, E.P.; Rybakov, N.Z.

    1977-01-01

    Construction of nuclear power plants in a highly populated region near large industrial centres necessitates to pay a special attention to their nuclear and radiation safety. Safety problems of nuclear reactor operation are discussed, in particular, they are: reliable stoppage of fission chain reaction at any emergency cases; reliable core cooling with failure of various equipment; emergency core cooling with breached pipes of a circulating circuit; and prevention of radioactive coolant release outside the nuclear power plant in amount exceeding the values adopted. Channel-type water boiling reactors incorporate specific features requiring a new approach to safety operation of a reactor and a nuclear power plant. These include primarily a rather large steam volume in the coolant circuit, large amount of accumulated heat, void reactivity coefficient. Channel-type reactors characterized by fair neutron balance and flexible fuel cycle, have a series of advantages alleviating the problem of ensuring their safety. The possibility of reliable control over the state of each channel allows to replace failed fuel elements by the new ones, when operating on-load, to increase the number of circulating loops and reduce the diameter of main pipelines, simplifies significantly the problem of channel emergency cooling and localization of a radioactive coolant release from a breached circuit. The concept of channel-type reactors is based on the solution of three main problems. First, plant safety should be assured in emergency switch off of separate units and, if possible, energy conditions should be maintained, this is of particular importance considering the increase in unit power. Second, the system of safety and emergency cooling should eliminate a great many failures of fuel elements in case of potential breaches of any tube in the circulating circuit. Finally, rugged boxes and localizing devices should be provided to exclude damage of structural elements of the nuclear power

  2. New graphite/salt materials for high temperature energy storage. Phase change properties study; Nouveaux materiaux graphite/sel pour le stockage d'energie a haute temperature. Etude des proprietes de changement de phase

    Energy Technology Data Exchange (ETDEWEB)

    Lopez, J

    2007-07-15

    This work is a contribution to the study of new graphite/salt composites dedicated to high temperature energy storage ({>=}200 C). The aim is to analyse and to understand the influence of both graphite and composite microstructure on the phase change properties of salts. This PhD is carried out within the framework of two projects: DISTOR (European) and HTPSTOCK (French). The major contributions of this work are threefold: 1) An important database (solid-liquid phase change properties) is provided from the DSC analysis of six salts and the corresponding composites. 2) Rigorous modeling of salts melting in confined media in several geometries are proposed to understand why, during the first melting of the compression elaborated composites, problems of salt leakage are observed. These models show that the materials morphology is responsible for these phenomena: the graphite matrix restrains the volume expansion due to salt melting: salt melts under pressure, which leads to a melting on a large temperature range and to a loss of energy density. Sensitivity analysis of parameters (geometric and physic) shows that matrix rigidity modulus is the parameter on which it is necessary to act during the composites elaboration to blur this phenomenon. 3) Finally, this work proposes a thermodynamic formulation of both surface/interface phenomena and the presence of dissolved impurities being able to explain a melting point lowering. It seems that the melting point lowering observed ({approx} 5 C) are mainly due to the presence of dissolved impurities (brought by graphite) in the liquid, along with an additional Gibbs-Thomson effect ({approx} 1 C, related to the size of the clusters crystals). (author)

  3. Mechanical properties of nuclear waste glasses

    International Nuclear Information System (INIS)

    Connelly, A.J.; Hand, R.J.; Bingham, P.A.; Hyatt, N.C.

    2011-01-01

    The mechanical properties of nuclear waste glasses are important as they will determine the degree of cracking that may occur either on cooling or following a handling accident. Recent interest in the vitrification of intermediate level radioactive waste (ILW) as well as high level radioactive waste (HLW) has led to the development of new waste glass compositions that have not previously been characterised. Therefore the mechanical properties, including Young's modulus, Poisson's ratio, hardness, indentation fracture toughness and brittleness of a series of glasses designed to safely incorporate wet ILW have been investigated. The results are presented and compared with the equivalent properties of an inactive simulant of the current UK HLW glass and other nuclear waste glasses from the literature. The higher density glasses tend to have slightly lower hardness and indentation fracture toughness values and slightly higher brittleness values, however, it is shown that the variations in mechanical properties between these different glasses are limited, are well within the range of published values for nuclear waste glasses, and that the surveyed data for all radioactive waste glasses fall within relatively narrow range.

  4. NMR studies on graphite-methanol system

    International Nuclear Information System (INIS)

    El-Akkad, T.M.

    1977-01-01

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

  5. FIRE-RESISTANCE PROPERTIES RESEARCH OF “WATER GLASS - GRAPHITE MICROPARTICLES” COMPOSITE MATERIAL

    Directory of Open Access Journals (Sweden)

    E. A. Pitukhin

    2016-03-01

    Full Text Available Subject of Research. Research results of the fire-resistance for “water glass - graphite microparticles” composite material (CM are given. The method for fire-resistance test of the micro composition is suggested in order to determine the limit state of the experimental samples under hightemperature action. Method. Test-benchequipment being used for research includes metering devices of temperature and time, as well as laboratory electric furnace PL20 with a maximum temperature in the chamber up to 1250ºC. Fire-resistance limit for the test samples of composite material is determined by the loss of insulating ability (I. For that purpose, the time is obtained from the test beginning with the standard temperature mode up to a limiting condition. Main Results. In accordance with the requirements of regulatory documents fire-resistance limit I15 has been obtained equal to 15 minutes. The qualitative and quantitative phase analysis of the CM structure has been done. By the study of samples by X-ray diffraction and electron microscopy we have determined that the material retains the same chemical structure with a monotonic heating above 700° C. Practical Relevance. The composite material with obtained characteristics can be used as a protective coating for building constructions with the aim of fire-resistance enhancement and fuel hazard reduction.

  6. Influence of atmospheric plasma on physicochemical properties of vapor-grown graphite nanofibers.

    Science.gov (United States)

    Seo, Min-Kang; Park, Soo-Jin; Lee, Sang-Kwan

    2005-05-01

    Vapor-grown graphite nanofibers (GNFs) were modified by plasma treatments using low-pressure plasmas with different gases (Ar gas only and/or Ar/O2 gases), flow rates, pressures, and powers. Surface characterizations and morphologies of the GNFs after plasma treatment were investigated by X-ray photoelectron spectroscopy (XPS), contact angle, titration, and transmission electron microscopy (TEM) measurements. Also, the investigation of thermomechanical behavior and impact strengths of the GNFs/epoxy composites was performed by dynamic-mechanical thermal analysis (DMTA) and Izod impact testing, respectively. The plasma treatment of the fibers changed the surface morphologies by forming a layer with a thickness on the order of 1 nm, mainly consisting of oxygen functional groups such as hydroxyl, carbonyl, and carboxyl groups. After functionalization of the complete surfaces, further plasma treatment did not enhance the superficial oxygen content but slightly changed the portions of the functional groups. Also, the composites with plasma-treated GNFs showed an increase in T(g) and impact strength compared to the composites containing the same amount of plasma-untreated GNFs.

  7. 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.

  8. Mechanical and Electrical Properties of Styrene-Isoprene-Styrene Copolymer Doped with Expanded Graphite Nanoplatelets

    Directory of Open Access Journals (Sweden)

    Zdenko Špitalský

    2015-01-01

    Full Text Available The molecular dynamics of a triblock copolymer and of expanded graphite nanoplatelets were investigated. Composites were prepared using the solution technique. The effects of filler addition and of filler-matrix interactions were investigated using dielectric relaxation spectroscopy (DRS and dynamic mechanical analysis (DMA. Only one relaxation was observed by DRS, which was associated with the relaxation of the main polymer chain. Both DRS and DMA demonstrated that the addition of the filler does not cause a significant change in either the temperature of the relaxation or its activation energy, which suggests the presence of weak interactions between the filler and matrix. The storage modulus of the composites increased with increasing filler content. The composite containing 8% filler exhibited a storage modulus increase of approximately 394% in the rubber area. Using the DC electrical conductivity measurements, the electrical percolation threshold was determined to be approximately 5%. The dielectric permittivity and conductivity in the microwave region were determined, confirming that percolating behavior and the critical threshold concentration.

  9. Graphite Microstructural Characterization Using Time-Domain and Correlation-Based Ultrasonics

    Energy Technology Data Exchange (ETDEWEB)

    Spicer, James [Johns Hopkins Univ., Baltimore, MD (United States)

    2017-12-06

    Among techniques that have been used to determine elastic modulus in nuclear graphites, ultrasonic methods have enjoyed wide use and standards using contacting piezoelectric tranducers have been developed to ensure repeatability of these types of measurements. However, the use of couplants and the pressures used to effectively couple transducers to samples can bias measurements and produce results that are not wholly related to the properties of the graphite itself. In this work, we have investigated the use of laser ultrasonic methods for making elastic modulus measurements in nuclear graphites. These methods use laser-based transmitters and receivers to gather data and do not require use of ultrasonic couplants or mechanical contact with the sample. As a result, information directly related to the elastic responses of graphite can be gathered even if the graphite is porous, brittle and compliant. In particular, we have demonstrated the use of laser ultrasonics for the determination of both Young’s modulus and shear modulus in a range of nuclear graphites including those that are being considered for use in future nuclear reactors. These results have been analyzed to assess the contributions of porosity and microcracking to the elastic responses of these graphites. Laser-based methods have also been used to assess the moduli of NBG-18 and IG-110 where samples of each grade were oxidized to produce specific changes in porosity. These data were used to develop new models for the elastic responses of nuclear graphites and these models have been used to infer specific changes in graphite microstructure that occur during oxidation that affect elastic modulus. Specifically, we show how ultrasonic measurements in oxidized graphites are consistent with nano/microscale oxidation processes where basal plane edges react more readily than basal plane surfaces. We have also shown the use of laser-based methods to perform shear-wave birefringence measurements and have shown

  10. 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.

  11. Staging properties of potassium-ammonia ternary graphite intercalation compounds at high ammonia pressure

    Science.gov (United States)

    Qian, X. W.; Solin, S. A.

    1989-04-01

    The pressure dependence of the (00l) x-ray diffraction patterns of the ternary graphite intercalation compound K(NH3)xC24 has been studied in the range 0.5-11 kbar (for which x~4.5) using a diamond anvil cell. A special apparatus for loading the cell with liquid ammonia at room temperature has been constructed and is briefly described. In these experiments, the pressure-transmitting fluid was also an intercalant, namely ammonia. Therefore, the chemical potential of this species was linearly coupled to the applied pressure in contrast to the usual case where the pressure-transmitting fluid is chemically passive. The pressure dependences of the basal spacings and of the relative intensities of key reflections have been measured, as have the compressibilities of the stage-1 and stage-2 components of the two-phase system. Basal-spacing anomalies and anomalies in the relative intensities occur at pressures of ~3.5 and 8.0 kbar and are tentatively attributed to in-plane coordination changes in the potassium-ammonia ratio. Using thermodynamic arguments and Le Chatelier's principle we show quantitatively that a staging phase transition from pure stage-1 phase to an admixture of stage-1 and stage-2 is expected with increased pressure above 10 bar in agreement with experiment. The saturation ammonia compositions (x values) of the admixed stages are found to be 4.5 and 5.4 for the stage-1 and -2 components, respectively. This result is interpreted as evidence that the composition is not sterically limited but is determined by the binding energy of ammonia for potassium and by the perturbation to this energy from the guest-host interaction.

  12. A Nanotube Surface Reinforced Graphite Fiber Exhibiting Significantly Enhanced Properties, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Nanotechnology which includes carbon nanotubes has the potential to produce materials that exhibit properties beyond those expected from conventional materials which...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  14. Graphite nanoplatelets-modified PLA/PCL: Effect of blend ratio and nanofiller localization on structure and properties.

    Science.gov (United States)

    Kelnar, Ivan; Kratochvíl, Jaroslav; Kaprálková, Ludmila; Zhigunov, Alexander; Nevoralová, Martina

    2017-07-01

    Structure and properties of poly(lactic acid) (PLA)/poly (ɛ-caprolactone) (PCL) influenced by graphite nanoplatelets (GNP) were studied in dependence on blend composition. Electron microscopy indicates predominant localization of GNP in PCL. GNP-induced changes in viscosity hinder refinement of PCL inclusions, support PCL continuity in the co-continuous system, and lead to reduction of PLA inclusions size without GNP being present at the interface in the PCL-matrix blend. Negligible differences in crystallinity of both phases indicate that mechanical behaviour is mainly influenced by reinforcement and GNP-induced changes in morphology. Addition of 5 parts of GNP leads to ~40% and ~25% increase of stiffness in the PCL- and PLA-matrix systems, respectively, whereas the reinforcing effect is practically eliminated in the co-continuous systems due to GNP-induced lower continuity of PLA which enhances toughness. Impact resistance of the 80/20 blend shows increase with 5 parts content due to synergistic effect of PCL/GNP stacks, whereas minor increase in the blend of the ductile PCL matrix with brittle PLA inclusions is caused by GNP-modification of the component parameters. Results indicate high potential of GNP in preparing biocompatible systems with wide range of structure and properties. Copyright © 2017 Elsevier Ltd. All rights reserved.

  15. Effects of resin content and preparing conditions on the properties of polyphenylene sulfide resin/graphite composite for bipolar plate

    Energy Technology Data Exchange (ETDEWEB)

    Xia, Li-gang; Li, Ai-ju; Yin, Qiang [Key Laboratory for Liquid Structure and Heredity of Materials, Ministry of Education, Shandong University, Shandong Key Laboratory of Engineering Ceramics, Shandong University, Jinan 250061 (China); Wang, Wei-qiang [School of Mechanical Engineering, Shandong University, Jinan 250061 (China); Lin, Heng; Zhao, Yi-bo [School of Material Science and Engineering, Shandong University, Jinan 250061 (China)

    2008-03-15

    In the paper, a kind of polyphenylene sulfide (PPS) resin/graphite (G) composite for bipolar plate was prepared by using the PPS resin as adhesive and simple hot pressing. The influences of the resin content, the molding temperature and holding time on the conductivity and the bending strength of the PPS/G composite bipolar plate were investigated firstly and then the optimum content and the preparing conditions of the composite were obtained. The experimental results show that the electrical conductivity decreases and the bending strength reveals a serrated variation with increase in PPS resin content; when the holding time is certain, the conductivity decreases and the bending strength increases with the molding temperature increasing. The experimental results further show that the effect of the holding time on the properties of the composite is different at different molding temperatures. The PPS/G composite with 20% PPS resin content has electrical conductivity of 118.9 S cm{sup -1} and bending strength of 52.4 MPa when it molded at 380 C for 30 min, and has electrical conductivity of 105 S cm{sup -1}, bending strength of 55.7 MPa when it molded at 390 C for 30 min. The properties of the composites can meet the requirements of United States Department of Energy (DOE). (author)

  16. Preparation and Thermoelectric Properties of Graphite/Bi0.5Sb1.5Te3 Composites

    Science.gov (United States)

    Hu, Wenhua; Zhou, Hongyu; Mu, Xin; He, Danqi; Ji, Pengxia; Hou, Weikang; Wei, Ping; Zhu, Wanting; Nie, Xiaolei; Zhao, Wenyu

    2018-06-01

    Bismuth telluride zone-melting alloys are the most commercially used thermoelectric materials. However, the zone-melting ingots have weak machinability due to the strong preferred orientation. Here, non-textured graphite/Bi0.5Sb1.5Te3 (G/BST) composites were prepared by a powder metallurgy method combined with cold-pressing and annealing treatments. The composition, microstructure, and thermoelectric properties of the G/BST composites with different mass percentages of G were investigated. It was found that G addition could effectively reduce the thermal conductivity and slightly improve the electrical properties of the BST, which resulted in a large enhancement in the figure-of-merit, ZT. The largest ZT for the xG/BST composites with x = 0.05% reached 1.05 at 320 K, which is increased by 35% as compared with that of the G-free BST materials. This work provided an effective method for preparing non-textured Bi2Te3-based TE materials with a simple process, low cost, and large potential in scale production.

  17. Frequency dependence of electrical properties of polyvinylidene fluoride/graphite electrode waste/natural carbon black composite

    Science.gov (United States)

    Insiyanda, D. R.; Indayaningsih, N.; Prihandoko, B.; Subhan, A.; Khaerudini, D. S.; Widodo, H.; Destyorini, F.; Chaer, A.

    2018-03-01

    Polyvinylidene fluoride (PVdF) is a semi-crystalline thermoplastic material with remarkably high piezoelectric coefficient and an attractive polymer matrix for micro-composite with superior mechanical and electrical properties. The conductive filler is obtained from Graphite Electrode Waste (GEW) and Natural Carbon Black (NCB). The variation of composite content (%) of PVdF/NCB/GEW were 100/0/0, 95/5/0, 95/0/5, 95/2.5/2.5. This experiment employed dry dispersion method for material mixing. The materials were then moulded using hot press machine with compression parameters of P = 5.5 MPa, T = 150 °C, t = 60 minutes, A = 5×5×(0.2 - 0.4) cm3. The electrical conductivity properties of pure PVdF, as well as PVdF/GEW, PVdF/NCB, and PVdF/NCB/GEW composites were investigated in a frequency range of 100 to 100000 Hz. The PVdF/GEW sample obtained the highest electrical conductivity. It is concluded that GEW and NCB can be incorporated into PVdF as a conductive filler to increase the conductivity of conductive material composite without solvent.

  18. 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...

  19. Irradiation Creep in Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Ubic, Rick; Butt, Darryl; Windes, William

    2014-03-13

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

  20. Path dependent models to predict property changes in graphite irradiated at changing irradiation temperatures

    CSIR Research Space (South Africa)

    Kok, S

    2010-10-01

    Full Text Available Property changes occur in materials subjected to irradiation. The bulk of experimental data and associated empirical models are for isothermal irradiation. The form that these isothermal models take is usually closed form expressions in terms...

  1. Polyphase diffusion of fission products in graphite

    International Nuclear Information System (INIS)

    Dannert, V.

    1989-05-01

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

  2. Asymmetric nuclear matter and neutron star properties

    International Nuclear Information System (INIS)

    Engvik, L.; Hjorth-Jensen, M.; Osnes, E.; Bao, G.; Oestgaard, E.

    1994-06-01

    Properties of neutron stars such as mass and radius, using a relativistic Dirac-Brueckner-Hartree-Fock approach, are calculated. Modern meson-exchange potential models are used to evaluate the G-matrix for asymmetric nuclear matter. For pure neutron matter the maximum mass is found to be M max ∼ 2.4M for a radius R ∼ 12 km. With a proton fraction of 30% the result is M max ∼ 2.1M for a radius R ∼ 10.5 km, close to the experimental values. The implications are discussed. 20 refs., 3 figs

  3. 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

  4. Glass-Graphite Composite Materials

    International Nuclear Information System (INIS)

    Mayzan, M.Z.H.; Lloyd, J.W.; Heath, P.G.; Stennett, M.C.; Hyatt, N.C.; Hand, R.J.

    2016-01-01

    A summary is presented of investigations into the potential of producing glass-composite materials for the immobilisation of graphite or other carbonaceous materials arising from nuclear power generation. The methods are primarily based on the production of base glasses which are subsequently sintered with powdered graphite or simulant TRISO particles. Consideration is also given to the direct preparation of glass-graphite composite materials using microwave technology. Production of dense composite wasteforms with TRISO particles was more successful than with powdered graphite, as wasteforms containing larger amounts of graphite were resistant to densification and the glasses tried did not penetrate the pores under the pressureless conditions used. Based on the results obtained it is concluded that the production of dense glassgraphite composite wasteforms will require the application of pressure. (author)

  5. The optical properties and photocatalytic activity of CdS-ZnS-TiO{sub 2}/Graphite for isopropanol degradation under visible light irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Rahmawati, Fitria, E-mail: fitria@mipa.uns.ac.id; Wulandari, Rini, E-mail: riniwulandari55@yahoo.com; Murni, Irvinna M., E-mail: irvinna-mutiara@yahoo.com; Mudjijono, E-mail: mbahparto@yahoo.com [Research Group of Solid State Chemistry & Catalysis, Chemistry Department, Sebelas Maret University, Jl. Ir. Sutami 36 A Kentingan, Surakarta, 57126 (Indonesia)

    2016-02-08

    This research prepared a photocatalyst tablet of CdS-ZnS-TiO{sub 2} on a graphite substrate. The synthesis was conducted through chemical bath deposition method. The graphite substrate used was a waste graphite rod from primary batteries. The aims of this research are studying the crystal structure, the optical properties and the photocatalytic activity of the prepared material. The photocatalytic activity was determined through isopropanol degradation. The result shows that the TiO{sub 2}/Graphite provide direct transition gap energy at 2.91 eV and an indirect transition gap energy at 3.21 eV. Deposition of CdS-ZnS changed the direct transition gap energy to 3.01 eV and the indirect transition gap energy to 3.22 eV. Isopropanol degradation with the prepared catalyst produced new peaks at 223-224 nm and 265-266 nm confirming the production of acetone. The degradation follows first order with rate constant of 2.4 × 10{sup −2} min{sup −1}.

  6. Effect of load, area of contact, and contact stress on the tribological properties of polyimide bonded graphite fluoride films

    Science.gov (United States)

    Fusaro, R. L.

    1981-01-01

    A pin-on-disk type of friction and wear apparatus was used to study the effect of load, contact stress and rider area of contact on the friction and wear properties of polyimide-bonded graphite fluoride films. Different rider area contacts were obtained by initially generating flats (with areas of 0.0035, 0.0071, 0.0145, and 0.0240 cm) on 0.476-cm radius hemispherically-tipped riders. Different projected contact stresses were obtained by applying loads of 2.5-to 58.8-N to the flats. Two film wear mechanisms were observed. The first was found to be a linear function of contact stress and was independent of rider area of contact. The second was found to increase exponentially as the stress increased. The second also appeared to be a function of rider contact area. Wear equations for each mechanism were empirically derived from the experimental data. In general, friction coefficients increased with increasing rider contact area and with sliding duration. This was related to the build-up of thick rider transfer films.

  7. Effect of graphite content on magnetic and mechanical properties of TiC-TiN-Mo-Ni cermets

    Science.gov (United States)

    Zhang, Man; Yang, Qingqing; Xiong, Weihao; Huang, Bin; Ruan, Linji; Mao, Qiao; Li, Shengtao

    2018-04-01

    TiC-10TiN-6Mo-xGr-yNi (mol%, Gr represents graphite, x = 0, 2, 4, 6, 8, and y = 15, 30) cermets were prepared by powder metallurgy method, in order to inverstigate the effect of Gr content on magnetic and mechanical properties of TiC-TiN-Mo-Ni cermets. Room-temperature (RT) saturation magnetization (Ms) and remanence (Mr) of cermets increased with increasing x. This was mainly attributed to that the total content of non-ferromagnetic carbonitride-forming elements Ti and Mo in Ni-based binder phase decreased with increasing x. At the same x, cermets for y = 15 had lower RT Ms and Mr than those for y = 30. Cermets containing more than 2 mol% Gr became ferromagnetic at RT. Bending strength of cermets first increased and then decreased with increasing x. It reached the maximum at x = 2, mainly due to high total content of solutes Ti and Mo in Ni-based binder phase, and moderate thickness of outer rim of Ti(C,N) ceramic grains. Hardness of cermets was not significantly affected by x, mainly due to the combined action of the decrease of the total content of Ti and Mo in binder phase and the increase of the volume fraction of ceramic grains. At the same x, cermets for y = 15 had lower bending strength and higher hardness than those for y = 30.

  8. Preparation and dielectric properties of novel composites based on oxidized styrene-butadienestyrene copolymer and polyaniline modified exfoliated graphite nanoplates

    Science.gov (United States)

    Lv, Qun-Chen; Li, Ying; Zhong, Zhi-Kui; Wu, Hui-Jun; He, Fu-An; Lam, Kwok-Ho

    2018-05-01

    To improve the dielectric performance of high-dielectric-constant conductive filler/polymer composites, polyaniline was deposited on exfoliated graphite nanoplates (xGNPs) by in-situ polymerization method to form polyaniline (PANI) coated xGNPs (xGNPs@PANI) as the conductive filler for the oxidized styrene-butadienestyrene copolymer (SBS-FH) containing both hydroxyl and formyloxy groups. The results of TEM, SEM, FTIR, TGA, Raman spectrum, XPS, and WAXD showed that PANI had been coated onto the surface of xGNPs successfully. The xGNPs@PANI/SBS-FH composites were prepared by a simple solution-blending method and the homogenous distribution of xGNPs@PANI in the SBS-FH matrix was confirmed by SEM. The presence of xGNPs@PANI was found to significantly improve the dielectric properties of resultant composite compared to the unmodified xGNPs. For example, the xGNPs@PANI/SBS-FH composite near percolation threshold filled with 9.38 vol.% xGNPs@PANI showed a dielectric constant of 56.8 and a dielectric loss factor of 0.51 at 1000 Hz, while the corresponding values of xGNPs (1.19 vol.%)/SBS composite were 15.96 and 2.91 at 1000 Hz, respectively. In addition, the incorporation of xGNPs@PANI into SBS-FH could effectively enhance the thermal conductivity of resultant xGNPs@PANI/SBS-FH composite.

  9. Effect of medium on friction and wear properties of compacted graphite cast iron processed by biomimetic coupling laser remelting process

    International Nuclear Information System (INIS)

    Guo Qingchun; Zhou Hong; Wang Chengtao; Zhang Wei; Lin Pengyu; Sun Na; Ren Luquan

    2009-01-01

    Stimulated by the cuticles of soil animals, an attempt to improve the wear resistance of compact graphite cast iron (CGI) with biomimetic units on the surface was made by using a biomimetic coupled laser remelting process in air and various thicknesses water film, respectively. The microstructures of biomimetic units were examined by scanning electron microscope and X-ray diffraction was used to describe the microstructure and identify the phases in the melted zone. Microhardness was measured and the wear behaviors of biomimetic specimens as functions of different mediums as well as various water film thicknesses were investigated under dry sliding condition, respectively. The results indicated that the microstructure zones in the biomimetic specimens processed with water film are refined compared with that processed in air and had better wear resistance increased by 60%, the microhardness of biomimetic units has been improved significantly. The application of water film provided finer microstructures and much more regular grain shape in biomimetic units, which played a key role in improving the friction properties and wear resistance of CGI.

  10. Effect of medium on friction and wear properties of compacted graphite cast iron processed by biomimetic coupling laser remelting process

    Science.gov (United States)

    Guo, Qing-chun; Zhou, Hong; Wang, Cheng-tao; Zhang, Wei; Lin, Peng-yu; Sun, Na; Ren, Luquan

    2009-04-01

    Stimulated by the cuticles of soil animals, an attempt to improve the wear resistance of compact graphite cast iron (CGI) with biomimetic units on the surface was made by using a biomimetic coupled laser remelting process in air and various thicknesses water film, respectively. The microstructures of biomimetic units were examined by scanning electron microscope and X-ray diffraction was used to describe the microstructure and identify the phases in the melted zone. Microhardness was measured and the wear behaviors of biomimetic specimens as functions of different mediums as well as various water film thicknesses were investigated under dry sliding condition, respectively. The results indicated that the microstructure zones in the biomimetic specimens processed with water film are refined compared with that processed in air and had better wear resistance increased by 60%, the microhardness of biomimetic units has been improved significantly. The application of water film provided finer microstructures and much more regular grain shape in biomimetic units, which played a key role in improving the friction properties and wear resistance of CGI.

  11. 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)

  12. Engineering properties of nuclear waste slurries - 16378

    International Nuclear Information System (INIS)

    Biggs, Simon; Fairweather, Michael; Hunter, Timothy; Omokanye, Qanitalillahi; Peakall, Jeffrey

    2009-01-01

    The type of particulate systems encountered in legacy nuclear waste slurries is highly complicated, with the aggregation and flow behaviour being at times very variable. However, deconstructing the complex overall slurry activity to singular particle-particle interactions can lead to a greater understanding of the mechanisms involved with particle aggregation, and so to predictions of their settling and flow in nuclear systems. Of particular importance to legacy waste is the role of salts in controlling the attraction of particles (and so in dictating the rheological properties of the system) as sludge may contain a variety of specific ions and generally have high ionic conductivity [1]. In this paper, particle-particle interactions are characterised using a number of complimentary methods, and their influence on resulting flow and bed compression is measured. The methods used to characterise the particle-particle interactions under various salt and pH conditions were electro-acoustic analysis (zeta potential) and atomic force microscopy (AFM). Following on from the analysis of particle-particle properties, bulk sediment behaviour was investigated using shear and compressive yield stress measurements, vital parameters in dictating flow and dewatering performance, respectively. Together, these techniques enable the characterisation of a range of particulate systems that may be encountered in legacy wastes, and results point to a number of important factors that can help explain the observed variability in industrial slurry behaviour. (authors)

  13. Mixed graphite cast iron for automotive exhaust component applications

    Directory of Open Access Journals (Sweden)

    De-lin Li

    2017-11-01

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

  14. Characterisation of Chlorine Behavior in French Graphite

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  15. 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)

  16. Development and properties of aluminum-clad graphite/epoxy tubes for space structures

    Science.gov (United States)

    Johnson, R. R.; Kural, M. H.

    1988-01-01

    This paper presents the development and properties of seamless aluminum-clad P75/Epoxy tubes and the unique manufacturing method used in their production. Thermo-mechanical properties of the tubes were determined analytically and verified by tests. These properties were shown to be suitable for space structures that require high stiffness, low weight and thermal expansion, and dimensional stability during operational life. A special feature of the tubes is the ability to tune the tube for thermal expansion after fabrication by a chemical milling process. The tubes are also resistant to atomic oxygen and handling damage. The toughness of the tubes was demonstrated by impact testing. Cyclic thermal testing showed no adverse effects on the expansion and stiffness behavior of the tubes. The paper also includes a discussion of a joining method that uses aluminum end fittings and an efficient scarf joint configuration. Additional studies considered various adhesives and fitting materials. Joint allowables were higher for titanium and B4C particulate magnesium fittings. The effect of different adhesives under static loading conditions favored the high-strength adhesives.

  17. Effect of High Energy Radiation on Mechanical Properties of Graphite Fiber Reinforced Composites. M.S. Thesis

    Science.gov (United States)

    Naranong, N.

    1980-01-01

    The flexural strength and average modulus of graphite fiber reinforced composites were tested before and after exposure to 0.5 Mev electron radiation and 1.33 Mev gamma radiation by using a three point bending test (ASTM D-790). The irradiation was conducted on vacuum treated samples. Graphite fiber/epoxy (T300/5208), graphite fiber/polyimide (C6000/PMR 15) and graphite fiber/polysulfone (C6000/P1700) composites after being irradiated with 0.5 Mev electron radiation in vacuum up to 5000 Mrad, show increases in stress and modulus of approximately 12% compared with the controls. Graphite fiber/epoxy (T300/5208 and AS/3501-6), after being irradiated with 1.33 Mev gamma radiation up to 360 Mrads, show increases in stress and modulus of approximately 6% at 167 Mrad compared with the controls. Results suggest that the graphite fiber composites studied should withstand the high energy radiation in a space environment for a considerable time, e.g., over 30 years.

  18. Management of UKAEA graphite liabilities

    International Nuclear Information System (INIS)

    Wise, M.

    2001-01-01

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

  19. Development of a pencil-type single shield graphite quasi-adiabatic calorimeter and comparison of its performance with a double-shield graphite calorimeter for the measurement of nuclear heat deposition rate in a fusion environment

    International Nuclear Information System (INIS)

    Joneja, O.P.; Rosselet, M.; Ligou, J.; Gardel, P.

    1995-01-01

    Recently, heat deposition rate measurements were reported that used a quasi-adiabatic double-shield graphite calorimeter. It was found that for a better understanding of nuclear heating due to incident radiation, having a calorimeter that could be conveniently moved axially and radially inside large material blocks would be advisable. Here, a simpler design, based on three elements, i.e., core, jacket, and shield is conceived. The fabrication and testing details are presented, and the performance of the current calorimeter is compared with a double-shield calorimeter under similar conditions. Such a system is found to be extremely sensitive and can be employed successfully at the LOTUS facility for future nuclear heat deposition rate measurements in large blocks of materials. The current design paves the way for the convenient testing of a large amount of kerma factor data required for constructing future fusion machines. The same configuration with minor changes can be extended to most of the fusion materials of interest. The core of the new calorimeter measures 11 mm in diameter and height and has overall dimensions of 24 mm in diameter and 180 mm in height. The response of the calorimeter is measured by placing it in front of the Haefely neutron generator. 12 refs., 16 figs., 9 tabs

  20. International nuclear safety center database on material properties

    International Nuclear Information System (INIS)

    Fink, J.K.

    1996-01-01

    International nuclear safety center database on the following material properties is described: fuel, cladding,absorbers, moderators, structural materials, coolants, concretes, liquid mixtures, uranium dioxide

  1. Purification and preparation of graphite oxide from natural graphite

    Energy Technology Data Exchange (ETDEWEB)

    Panatarani, C., E-mail: c.panatarani@phys.unpad.ac.id; Muthahhari, N.; Joni, I. Made [Instrumentation Systems and Functional Material Processing Laboratory, Department of Physics, Faculty of Mathematics and Natural Sciences, Universitas Padjadjaran, Padjadjaran University, Jl. Raya Bandung-Sumedang KM 21, Jatinangor, 45363, Jawa Barat (Indonesia); Rianto, Anton [Grafindo Nusantara Ltd., Belagio Mall Lantai 2, Unit 0 L3-19, Kawasan Mega Kuningan, Kav. B4 No.3, Jakarta Selatan (Indonesia)

    2016-03-11

    Graphite oxide has attracted much interest as a possible route for preparation of natural graphite in the large-scale production and manipulation of graphene as a material with extraordinary electronic properties. Graphite oxide was prepared by modified Hummers method from purified natural graphite sample from West Kalimantan. We demonstrated that natural graphite is well-purified by acid leaching method. The purified graphite was proceed for intercalating process by modifying Hummers method. The modification is on the reaction time and temperature of the intercalation process. The materials used in the intercalating process are H{sub 2}SO{sub 4} and KMNO{sub 4}. The purified natural graphite is analyzed by carbon content based on Loss on Ignition test. The thermo gravimetricanalysis and the Fouriertransform infrared spectroscopy are performed to investigate the oxidation results of the obtained GO which is indicated by the existence of functional groups. In addition, the X-ray diffraction and energy dispersive X-ray spectroscopy are also applied to characterize respectively for the crystal structure and elemental analysis. The results confirmed that natural graphite samples with 68% carbon content was purified into 97.68 % carbon content. While the intercalation process formed a formation of functional groups in the obtained GO. The results show that the temperature and reaction times have improved the efficiency of the oxidation process. It is concluded that these method could be considered as an important route for large-scale production of graphene.

  2. Mechanical Properties and Wear Characteristics Al-ZrO2-SiCp and Graphite Hybrid Metal Matrix Composites

    Science.gov (United States)

    Nayak, S. K.; Mahanta, T.; Sahoo, J. K.; Mishra, A.

    2018-03-01

    Development of Aluminum Metal Matrix Co mposites (AMMCs) has been one of the major requirements in engineering applicat ions due to their excellent mechanical properties, light weight and high strength. In the present investigation, Stir casting technique has been used for fabrication of co mposites, taking Alu miniu m as parent metal, Silicon Carbide (SiCp) of 7 vol. % of 220 mesh size and 1.75 vol. % of graphite as reinforcements. The Zirconia content was varied as 2.75, 4.5 and 6 vol. % to fabricate three d ifferent types of hybrid composites. The tensile strength and hardness were measured in UTM and Vickers hardness tester respectively and the wear characteristics were studied in a pin on disc friction monitor under dry sliding condition against steel counter face. The tensile strength was found to be 90 MPa, 120 MPa, 130 MPa and hardness 80.25 VHN, 103.22 VHN, 103.77 VHN for 2.75, 4.5 and 6vol. % of Zirconia respectively. Fro m the above investigation, it is recommended that composition with Al, 7 %-SiCp, 1.75 % -Gr and 6 vol %-ZrO2 showed better mechanical p roperties i.e . h igh tensile strength (130MPa) and reasonably good hardness (103.77 VHN) . The co mposite with Al, 7 % - SiCp, 1.75 % -Gr and 6 %-ZrO2 is good for short run frictional applicat ion and the composite with Al, 7 %- SiCp, 1.75 % -Gr and 4.5 %- ZrO2 may be used for long run frictional applicat ions after testing.

  3. Graphene-based materials via benzidine-assisted exfoliation and reduction of graphite oxide and their electrochemical properties

    Science.gov (United States)

    Vermisoglou, E. C.; Giannakopoulou, T.; Romanos, G.; Boukos, N.; Psycharis, V.; Lei, C.; Lekakou, C.; Petridis, D.; Trapalis, C.

    2017-01-01

    Benzidine, a compound bearing aromatic rings and terminal amino groups, was employed for the intercalation and simultaneous reduction of graphite oxide (GO). The aromatic diamine can be intercalated into GO as follows: (1) by grafting with the epoxy groups of GO, (2) by hydrogen bonding with the oxygen containing groups of GO. Stacking between benzidine aromatic rings and unoxidized domains of GO may occur through π-π interaction. The role of benzidine is influenced by pH conditions and the weight ratio GO/benzidine. Two weight ratios were tested i.e. 1:2 and 1:3. Under strong alkaline conditions through K2CO3 addition (pH ∼10.4-10.6) both intercalation and reduction of GO via amino groups occur, while under strong acidic conditions through HCl addition (pH ∼1.4-2.2) π-π stacking is preferred. When no base or acid is added (pH ∼5.2) and the weight ratio is 1:2, there are indications that reduction and π-π stacking occur, while at a GO/benzidine weight ratio 1:3 intercalation via amino groups and reduction seem to dominate. The aforementioned remarks render benzidine a multifunctional tool towards production of reduced graphene oxide. The effect of pH conditions and the GO/benzidine weight ratio on the quality and the electrochemical properties of the produced graphene-based materials were investigated. Cyclic voltammetry measurements using three-electrode cell and KCl aqueous solution as an electrolyte gave specific capacitance values up to ∼178 F/g. When electric double-layer capacitors (EDLC) were fabricated from these materials, the maximum capacitance in organic electrolyte i.e., tetraethyl ammonium tetrafluoroborate (TEABF4) in polycarbonate (PC) was ∼29 F/g.

  4. Phase relations, crystal structures and physical properties of nuclear fuels

    International Nuclear Information System (INIS)

    Tagawa, Hiroaki; Fujino, Takeo; Tateno, Jun

    1975-07-01

    Phase relations, crystal structures and physical properties of the compounds for nuclear fuels are presented, including melting point, thermal expansion, diffusion and magnetic and electric properties. Emphasis is on oxides, carbides and nitrides of thorium, uranium and plutonium. (auth.)

  5. 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.

  6. Preparation and microwave absorbing properties of nickel-coated graphite nanosheet with pyrrole via in situ polymerization

    International Nuclear Information System (INIS)

    Yang Yongqing; Qi Shuhua; Wang Jianning

    2012-01-01

    Highlights: ► NanoG was prepared and chemically coated by nano-Ni. ► PPy/NanoG and PPy/Ni/NanoG were prepared via in situ polymerization. ► The samples were characterized by SEM, XRD, EDS and FTIR. ► PPy/NanoG and PPy/Ni/NanoG exhibit good electromagnetic properties. - Abstract: Nanocomposites PPy/NanoG and PPy/Ni/NanoG were prepared via in situ polymerization of pyrrole in the presence of NanoG and nickel-coated graphite nanosheet (Ni/NanoG), respectively. The morphologies and nanostructures of NanoG, Ni/NanoG, PPy, PPy/NanoG and PPy/Ni/NanoG were characterized by scanning electron microscope (SEM), energy dispersive spectroscopy (EDS), Fourier transmission infrared (FTIR) and X-ray diffraction analysis (XRD). Results show that most of PPy chains disperse on NanoG and Ni/NanoG's surfaces for the high aspect ratio (300–500) of NanoG and Ni/NanoG. From the thermogravimetric analysis (TG) it can be seen that the introduction of Ni and NanoG leads the composites PPy/NanoG and PPy/Ni/NanoG to exhibit a better thermal stability than pure PPy. According to the four-point-probe test, the conductivities of the final PPy/NanoG and PPy/Ni/NanoG composites are dramatically increased compared to pure PPy. Measurement of electromagnetic parameters shows that the reflection loss (R) of PPy/Ni/NanoG is below −19 dB at the X band (8.2–12.4 GHz) and the minimum loss value is −23.46 dB at 9.88 GHz. The reflection loss of PPy/NanoG is below −10 dB at 8.2–12.4 GHz and the minimum loss value is −13.44 dB at 10.28 GHz. The microwave absorbing properties of PPy/NanoG and PPy/Ni/NanoG are superior to those of PPy.

  7. Properties of solid electrolyte interphase formed by prop-1-ene-1,3-sultone on graphite anode of Li-ion batteries

    International Nuclear Information System (INIS)

    Li, Bin; Xu, Mengqing; Li, Benzhen; Liu, Yanlin; Yang, Liang; Li, Weishan; Hu, Shejun

    2013-01-01

    Highlights: • SEI formed by PES on NG was characterized with charge/discharge test, SEM, FTIR, and XPS. • NG in PC-based electrolyte can be well protected using PES. • Sulfur-containing species is the main component of the SEI formed by PES. • Preferable reduction of PES results in the formation of protective SEI on NG. -- Abstract: The physical and chemical properties of the solid electrolyte interphase (SEI) formed by prop-1-ene-1,3-sultone (PES) on graphite anode in propylene carbonate (PC) based electrolyte for lithium ion battery were investigated by charge–discharge test, scanning electron spectroscopy with energy dispersive X-ray spectroscopy (SEM–EDS), Fourier transform infrared spectroscopy (FTIR), and X-ray photoelectron spectroscopy (XPS). It is found that the charge–discharge performance of the cell LiCoO 2 /natural graphite (NG) using PC-based electrolyte containing 3 wt% PES is superior to that containing 6 wt% propane sultone (PS), an SEI formation additive that has the similar molecular structure to PES but is reduced not as easily as PES. The results from SEM–EDS, FTIR and XPS show that the structure of graphite has been protected and some S-containing species are proven to be components of the SEI, suggesting that the preferable reduction of PES plays an important role in the formation of a protective SEI on NG

  8. Gross properties of nuclei and nuclear excitations

    International Nuclear Information System (INIS)

    Feldmeier, H.

    1992-01-01

    Heavy ion reactions are the main topics of the workshop. The different aspects that are discussed are: nuclear matter at relativistic energies, the transformation to quark matter, particle production, spallation and theoretical models for nuclear reaction kinetics. (BBOE)

  9. Gross properties of nuclei and nuclear excitations

    International Nuclear Information System (INIS)

    Feldmeier, H.

    1989-01-01

    These proceedings contain the articles presented at the named workshop. These concern new developments in nuclear structure studies, relativistic heavy ion reactions, and advanced techniques for nuclear spectroscopy. (HSI)

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  11. Gross properties of nuclei and nuclear excitations

    International Nuclear Information System (INIS)

    Feldmeier, H.

    1991-01-01

    These proceedings contain the articles presented at the named workshop. They deal with nuclear multifragmentation, heavy ion reaction kinetics, breakup and transfer processes in heavy ion reactions, the production of hypernuclei, nuclear structure in the framework of the quark model and QCD, and particle production in nuclear reactions. (HSI)

  12. 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)

  13. Enhanced photocatalytic property of hybrid graphitic C3N4 and graphitic ZnO nanocomposite: the effects of interface and doping

    Science.gov (United States)

    Cui, Jie; Liang, Shuhua; Sun, Shaodong; Zheng, Xing; Zhang, Jianmin

    2018-05-01

    Using first-principles calculations, we present a potential new way to improve the photocatalytic efficiency of the g-C3N4 sheet by coupling with the g-ZnO sheet to form heterojunction nanostructure followed by the addition of N atom at an atomic level. The result indicates the g-C3N4/g-ZnO heterojunction is a staggered band alignment (type II) structure and a polarized field is generated by the electrons transfer across the interface simultaneously, which facilitate the separation of e‑–h+ pairs and promote the photocatalytic activity. Furthermore, a great difference in energy levels between redox potentials and band edges of the C3N4/g-ZnO nanocomposite ensures that the water splitting/CO2 reduction reaction is energetically favored. In addition, through the incorporation of nitrogen dopant, the g-C3N4/N-g-ZnO nanocomposite displays desirable properties. The N-derived doping peak causes a decrease of the band gap width of the g-C3N4/g-ZnO nanocomposite, resulting in the enhanced optical absorption from UV into visible light. This theoretical predictions provide insightful outlooks in understanding the effects of interface and doping on the enhanced photocatalytic property of the g-C3N4/g-ZnO nanocomposites, which will assist in engineering highly efficient g-C3N4-based photocatalysts.

  14. Effects of fast-neutron damage from 0 to 42 x 1021 neutrons/CM2 on the physical properties of near-isotropic grades of graphite

    International Nuclear Information System (INIS)

    Cook, W.H.; Kennedy, C.R.; Eatherly, W.P.

    1975-01-01

    The characterization of property changes in various grades of near-isotropic, ''binderless'' grades of graphite as functions of fluence accumulated at 715 0 C from 0 to 42 x 10 21 neutrons/cm 2 (E greater than 50 keV) was made. Generally, the average coefficients of thermal expansion (CTE) from 20 to 600 0 C and the room-temperature values for strengths, fracture strains, Young's moduli, shear moduli, and calculated figures of merit (FOM) for resistance to thermal shock all ultimately decreased with fluence. (U.S.)

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

    International Nuclear Information System (INIS)

    Hong, Jin Ki; Chi, Se Hwan

    2006-01-01

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

  16. Thermally exfoliated graphite oxide

    Science.gov (United States)

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

    2011-01-01

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

  17. Experimental evaluation of brazed molybdenum-graphite bonds for the divertor of the NET/ITER nuclear fusion device

    International Nuclear Information System (INIS)

    Smid, I.; Linke, J.; Nickel, H.; Kny, E.; Reheis, N.; Kneringer, G.; Bolt, H.

    1995-01-01

    Composites consisting of plasma-facing carbon material brazed to molybdenum (TZM) substrates are a promising system for the divertor of the Next European Torus (NET) and the International Thermonuclear Experimental Reactor (ITER). Isotropic graphite and a refractory metal (molybdenum or TZM, a high temperature alloy of molybdenum), two dissimilar substrate materials, yet closely matched in their thermal expansivities, were joined with the use of four different high-temperature brazes: Zr, 90Ni-10Ti, 90Cu- 10Ti, and 70Ag-27Cu-3Ti (compositions in wt%). A summary is given of experiments on mechanical strength, heat transfer capability, structural changes, and failure modes under high heat loads of brazed bonds. Tensile-strength tests on the brazing interface prove the suitability of the brazes up to their melting point. The expected enhancement in thermal contact compared with graphite is confirmed. Passively cooled tiles of dimensions 25 mm x 25 mm were subjected to thermal cycling in electron-beam simulations. Heat fluxes of up to 10 MW m -2 were applied. (author)

  18. Experimental evaluation of brazed molybdenum-graphite bonds for the divertor of the NET/ITER nuclear fusion device

    International Nuclear Information System (INIS)

    Smid, Ivica; Linke, Jochen; Nickel, Hubertus; Kny, Erich; Reheis, Nikolaus; Kneringer, Guenther; Bolt, Harald

    1990-01-01

    Composites consisting of plasma-facing carbon material brazed to molybdenum (TZM) substrates are a promising system for the divertor of the Next European Torus (NET) and the International Thermonuclear Experimental Reactor (ITER). Isotropic graphite and a refractory metal (molybdenum or TZM, a high temperature alloy of molybdenum), two dissimilar substrate materials, yet closely matched in their thermal expansivities, were joined with the use of four different high-temperature brazes: Zr,90Ni-10Ti,90Cu-10Ti, and 70Ag-27Cu-3Ti(compositions in wt%). A summary is given of experiments on mechanical strength, heat transfer capability, structural changes, and failure modes under high heat loads of brazed bonds. Tensile-strength tests on the brazing interface prove the suitability of the brazes up to their melting point. The expected enhancement in thermal contact compared with graphite is confirmed. Passively cooled tiles of dimensions 25 mm x 25 mm were subjected to thermal cycling in electron-beam simulations. Heat fluxes of up to 10 MW m -2 were applied. (author)

  19. Non Destructive Thermal Analysis and In Situ Investigation of Creep Mechanism of Graphite and Ceramic Composites using Phase-sensitive THz Imaging & Nonlinear Resonant Ultrasonic Spectroscopy

    International Nuclear Information System (INIS)

    Zhang, XI-Cheng; Redo-Scanchez, Albert

    2012-01-01

    In this project, we conducted a comprehensive study on nuclear graphite properties with terahertz (THz) imaging. Graphite samples from Idaho National Laboratory were carefully imaged by continuous wave (CW) THz. The CW THz imaging of graphite shows that the samples from different billet with different fabricating conditions have different pore size and structure. Based on this result, we then used a phase sensitive THz system to study the graphite properties. In this exploration, various graphite were studied. By imaging nuclear graphite samples in reflection mode at nine different incident polarization angles using THz time-domain spectroscopy, we find that different domain distributions and levels of porosity will introduce polarization dependence in THz reflectivity. Sample with higher density is less porous and has a smaller average domain distribution. As a consequence, it is less polarization-dependent and the polarization-dependent frequency is higher. The results also show that samples oxidized at higher temperatures tend to be more polarization dependent. The graphite from the external billet is more polarization dependent compared to that from the center billet. In addition, we performed laser-based ultrasonic measurements on these graphite samples. The denser, unoxidized samples allow surface acoustic waves to propagate more rapidly than in the samples that had already undergone oxidation. Therefore, for the oxidized samples, the denser samples show less polarization-dependence, higher polarization-dependent frequency, and allow the surface acoustic waves propagate faster.

  20. Investigation on wear behavior of graphite baII under different pneumatic conveying environments

    International Nuclear Information System (INIS)

    Chen Zhipeng; Zheng Yanhua; Shi Lei; Yu Suyuan

    2014-01-01

    An experimental platform was built in the Institute of Nuclear and New Energy Technology (INET) to investigate the wear behavior of the graphite ball under the operational condition of the high temperature gas-cooled reactor (HTGR) fuel handling system. In this experimental platform, a series of experiments were carried out under different pneumatic conveying environments with the graphite balls, which were made of the material same as the fuel element matrix graphite (A3) of the 10 MW high temperature gas cooled reactor (HTR-10). The effect of the pneumatic conveying condition on the wear rate of graphite ball has been investigated, and the results include: (1) There is an obvious linear relationship between the wear rate and the feeding velocity of graphite ball elevated in the stainless steel elevating tube, and the wear rate will increase with the increase of the feeding velocity. (2) The wear rate of graphite ball under helium environment is significantly greater than that under air and nitrogen environments, which is caused by the different effects of various gas environments on mechanical properties of graphite. (author)

  1. Graphite nanoplatelets-modified PLA/PCL: effect of blend ratio and nanofiller localization on structure and properties

    Czech Academy of Sciences Publication Activity Database

    Kelnar, Ivan; Kratochvíl, Jaroslav; Kaprálková, Ludmila; Zhigunov, Alexander; Nevoralová, Martina

    2017-01-01

    Roč. 71, July (2017), s. 271-278 ISSN 1751-6161 R&D Projects: GA ČR(CZ) GA16-03194S Institutional support: RVO:61389013 Keywords : poly (epsilon-caprolactone) * poly (lactic acid) * graphite nanoplatelets Subject RIV: JI - Composite Materials OBOR OECD: Composites (including laminates, reinforced plastics, cermets, combined natural and synthetic fibre fabrics Impact factor: 3.110, year: 2016

  2. Gross properties of nuclei and nuclear excitations

    International Nuclear Information System (INIS)

    Feldmeier, H.

    1986-01-01

    These proceedings contain the invited and contributed papers presented at the named workshop. They deal with special topics in high- and intermediate-energy nuclear physics. Especially considered are the quark-gluon plasma, ultrarelativistic heavy ion reactions, and the application of QCD to nuclear structure calculations. (HSI)

  3. Gross properties of nuclei and nuclear excitations

    International Nuclear Information System (INIS)

    Feldmeier, H.

    1988-01-01

    These proceedings contain the articles presented at the named workshop. They are concerned with the application of the results from high-energy physics to nuclear structure calculations, thermodynamics of nuclear matter, and kinetics of heavy ion reactions with special regard to relativistic processes. (HSI)

  4. Bridged graphite oxide materials

    Science.gov (United States)

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

    2010-01-01

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

  5. Investigation of mono/competitive adsorption of environmentally relevant ionized weak acids on graphite: impact of molecular properties and thermodynamics.

    Science.gov (United States)

    Moustafa, Ahmed M A; McPhedran, Kerry N; Moreira, Jesús; Gamal El-Din, Mohamed

    2014-12-16

    The thermodynamics of adsorption and competitive interactions of five weak acids on a graphite surface was assessed in alkaline solutions. Adsorption of the acids in mono- and multicompound solutions followed their Freundlich isotherms which suggest a diversity of graphite adsorption sites as confirmed by the presence of carboxylic and phenolic groups observed on graphite surfaces. Thermodynamic calculations assigned the formation of the negatively charged assisted hydrogen bond (-CAHB) between ionized solutes and adsorbent surface groups as the possible adsorption mechanism. However, the similar pKa values of current acids resulted in comparable free energies for -CAHB formation (ΔG(-CAHB)) being less than solvation free energies (ΔGSolv). Thus, additional ΔG is supplemented by increased hydrophobicity due to proton exchange of ionized acids with water (ΔΔG Hydrophobicity). Adsorption capacities and competition coefficients indicated that ΔΔG Hydrophobicity values depend on the neutral and ionized acid Kow. Competitive adsorption implies that multilayer adsorption may occur via hydrophobic bonding with the CH3 ends of the self-assembled layer which affects the acid adsorption capacities in mixtures as compared to monocompound solutions. The determination of adsorption mechanisms will assist in understanding of the fate and bioavailability of emerging and classical weak acids released into natural waters.

  6. Attenuation of thermal neutron through graphite

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  7. 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

  8. 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

  9. Influence of the graphite type in the mechanical properties of graphene nanosheets and ultra high molecular weight polyethylene (GN/UHMWPE) nanocomposites

    International Nuclear Information System (INIS)

    Lima, M.E. de A.; Coutinho, S.V.; Fim, F. de C.

    2016-01-01

    This paper studies the mechanical properties of graphene nanosheets and ultra high molecular weight polyethylene (GN/UHMWPE) nanocomposites. The graphene nanosheets were obtained by two methods: the first, from Graphite Micrograf HC11 after 8 hours in a ultrasound bath in 70% alcoholic solution; the second, from Graflake 9950 prepared by oxidizing with H 2 SO 4 /HNO 3 (4:1), followed by thermal expansion at 1050°C for 30 seconds and ultrasound bath for 20 hours. Both were oven dried for 24 hours at 110°C and subsequently added to the polymer in percentages of 0.5, 1.0 and 1.5% (w/w). The powders of the nanocomposites were mixed in a ball mill for 1 hour and molded by hot compression. XRD and SEM images was possible to verify that the Graflake's GN had higher delamination that the Micrograph's GN. The mechanical properties were affected in different percentages for each graphite type. (author)

  10. Process for purifying graphite

    International Nuclear Information System (INIS)

    Clausius, R.A.

    1985-01-01

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

  11. On the thermal properties of polarized nuclear matter

    International Nuclear Information System (INIS)

    Hassan, M.Y.M.; Montasser, S.S.; Ramadan, S.

    1979-08-01

    The thermal properties of polarized nuclear matter are calculated using Skyrme III interaction modified by Dabrowski for polarized nuclear matter. The temperature dependence of the volume, isospin, spin and spin isospin pressure and energies are determined. The temperature, isospin, spin and spin isospin dependence of the equilibrium Fermi momentum is also discussed. (author)

  12. On the spin saturation and thermal properties of nuclear matter

    International Nuclear Information System (INIS)

    Hassan, M.Y.M.; Ramadan, S.

    1983-12-01

    The binding energy and the incompressibility of nuclear matter with degree of spin saturation D is calculated using the Skyrme interaction and two forms of a velocity dependent effective potential. The effect of the degree of spin saturation D on the thermal properties of nuclear matter is also discussed. It is found that generally the pressure decreases with increasing D. (author)

  13. Graphite selection for the FMIT test cell

    International Nuclear Information System (INIS)

    Morgan, W.C.

    1982-06-01

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

  14. Variability of mechanical properties of nuclear pressure vessel steels

    International Nuclear Information System (INIS)

    Petrequin, P.; Soulat, P.

    1980-01-01

    Causes of variability of mechanical properties nuclear pressure vessel steels are reviewed and discussed. The effects of product shape and size, processing history and heat treatment are investigated. Some quantitative informations are given on the scatter of mechanical properties of typical pressure vessel components. The necessity of using recommended or standardized properties for comparing mechanical properties before and after irradiation in pin pointed. (orig.) [de

  15. Data Report on the Newest Batch of PCEA Graphite for the VHTR Baseline Graphite Characterization Program

    Energy Technology Data Exchange (ETDEWEB)

    Carroll, Mark Christopher [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cottle, David Lynn [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rohrbaugh, David Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    This report details a comparison of mechanical and physical properties from the first billet of extruded PCEA nuclear-grade graphite from the newest batch of PCEA procured from GrafTech. Testing has largely been completed on three of the billets from the original batch of PCEA, with data distributions for those billets exhibiting a much wider range of values when compared to the distributions of properties from other grades. A higher propensity for extremely low values or specimens that broke while machining or handling was also characteristic of the billets from the first batch, owing to unusually large fissures or disparate flaws in the billets in an as-manufactured state. Coordination with GrafTech prior to placing the order for a second batch of PCEA included discussions on these large disparate flaws and how to prevent them during the manufacturing process. This report provides a comparison of the observed data distributions from properties measured in the first billet from the new batch of PCEA with those observed in the original batch, in order that an evaluation of tighter control of the manufacturing process and the outcome of these controls on final properties can be ascertained. Additionally, this billet of PCEA is the first billet to formally include measurements from two alternate test techniques that will become part of the Baseline Graphite Characterization database – the three-point bend test on sub-sized cylinders and the Brazilian disc splitting tensile strength test. As the program moves forward, property distributions from these two tests will be based on specimen geometries that match specimen geometries being used in the irradiated Advanced Graphite Creep (AGC) program. This will allow a more thorough evaluation of both the utility of the test and expected variability in properties when using those approaches on the constrained geometries of specimens irradiated in the Advanced Test Reactor as part of the AGC experiment.

  16. Effects of high temperature neutron irradiation on the physical, chemical and mechanical properties of fine-grained isotropic graphite

    International Nuclear Information System (INIS)

    Matsuo, H.; Nomura, S.; Imai, H.; Oku, T.; Eto, M.

    1987-01-01

    Effects of neutron irradiation on the dimensional change, coefficient of thermal expansion(CTE), thermal conductivity, corrosion rate, Young's modulus and strengths were studied for the candidate graphite material IG-110 of the experimental very high temperature gas-cooled reactor(VHTR) after irradiation at 585 - 1273 deg C to neutron fluences of up to about 3 x 10 25 n/m 2 (E > 29 fJ) in the JMTR and JRR-2, and to about 7 x 10 25 n/m 2 (E > 29 fJ) in the HFR. The results were compared with the irradiation behaviors of other graphites. Dimensional shrinkage was observed in the whole irradiation temperature range, showing lower value than 2 %. The shrinkage rate showed the minimum in the irradiation temperature of around 850 deg C, followed by the increase for the samples irradiated at higher temperatures. The dimensional stability of the material was clarified to be almost the same with that of H451 graphite. The CTE, thermal resistivity and Young's modulus increased in the early stage of irradiation and then only the CTE decreased while the thermal resistivity and Young's modulus levelled off with further irradiation. The neutron fluence showing the maximum CTE shifted to the lower fluence with increasing irradiation temperature. The increases of both thermal resistivity and Young's modulus were remarkable for the samples irradiated at lower temperatures. Compressive and bending strengths measured at room temperature increased after irradiation as well. The corrosion rate with water-vapor of 0.65 % in helium at high temperatures decreased owing to irradiation and the reduction was independent of irradiation temperature and neutron fluence. The activation energy for the reaction was estimated to be the same before and after irradiation. (author)

  17. Nuclear materials thermo-physical property database and property analysis using the database

    International Nuclear Information System (INIS)

    Jeong, Yeong Seok

    2002-02-01

    It is necessary that thermo-physical properties and understand of nuclear materials for evaluation and analysis to steady and accident states of commercial and research reactor. In this study, development of nuclear materials thermo-properties database and home page. In application of this database, it is analyzed of thermal conductivity, heat capacity, enthalpy, and linear thermal expansion of fuel and cladding material and compared thermo-properties model in nuclear fuel performance evaluation codes with experimental data in database. Results of compare thermo-property model of UO 2 fuel and cladding major performance evaluation code, both are similar

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

    International Nuclear Information System (INIS)

    Mohanty, Subhasish; Majumdar, Saurindranath; Srinivasan, Makuteswara

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-15

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

  20. Separation of Graphitic Line in Debyegram of the Reactor Graphite; Separation de la raie graphitique dans le debyegramme du graphite nucleaire; Otdelenie linii grafita v debaiegramme reaktornogo grafita; Separacion de la linea grafitica en el diagrama de debye del grafito nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Pandic, B. [Institut Rudjer Boskovio, Zagreb, Yugoslavia (Croatia)

    1963-11-15

    The author sets forth a method for the determination of crystallite dimensions L{sub c} in reactor graphite. The method consists of the mathematical correction of the diffraction profile (002) obtained by means of a Philips diffractometer with scintillation counter. During the work CuK{sub {beta}} radiation is used. The convenience of using CuK{sub {beta}} radiation rather than that of a CuK{sub {alpha}} doublet is discussed. All instrumental errors except that of the sample adsorption are eliminated using corresponding slits. Sample adsorption is corrected mathematically. Highly symmetrical lines of graphitic and non-graphitic phase could be analytically separated from such corrected line profiles. Experimental error in the determination of the half-height and half-width of these lines does not exceed 5%. (author) [French] On a elabore une methode pour determiner la hauteur moyenne L{sub c} des cristallites du graphite nucleaire. Il s'agit d'une methode mathematique de correction du profil de la raie de diffraction (002), obtenu par le diffractometre Philips avec un compteur a scintillation. On a discute les avantages des rayons CuK{sub {beta}} par rapport au doublet CuK{sub {alpha}}. Le choix des ouvertures elimine toutes les erreurs, sauf l'erreur d'absorption dans les echantillons, que l 'on ecarte mathematiquement. Les raies de diffraction pures et tres symetriques des phases graphitique et non graphitique peuvent etre separees du profil corrige de cette facon. L'incertitude dans la mesure de la demi-largeur de ces raies n'est pas superieure a 5%. (author) [Spanish] El autor de la memoria expone un metodo para determinar la altura media Lc de las cristalitas del grafito nuclear. Consiste en corregir matematicamente el perfil de la linea de difraccion (002) obtenido con un difractometro Philips y un contador de centelleo. En la memoria se explican las ventajas que la radiacion CuK{sub {beta}} ofrece sobre el doblete CuK{sub {alpha}} Si se elige correctamente la

  1. Fabrication of Al/Graphite/Al2O3 Surface Hybrid Nano Composite by Friction Stir Processing and Investigating The Wear and Microstructural Properties of The Composite

    Directory of Open Access Journals (Sweden)

    A. Mostafapour

    2012-10-01

    Full Text Available Friction stir processing was applied for fabricating an aluminum alloy based hybrid nano composite reinforced with nano sized Al2O3 and micro sized graphite particles. A mixture of Al2O3 and graphite particles was packed into a groove with 1 mm width and 4.5 mm depth, which had been cut in 5083 aluminum plate of 10 mm thick. Packed groove was subjected to friction stir processing in order to implement powder mixture into the aluminum alloy matrix. Microstructural properties were investigated by means of optical microscopy and scanning electron microscopy (SEM. It was found that reinforcement particle mixture was distributed uniformly in nugget zone. Wear resistance of composite was measured by dry sliding wear test. As a result, hybrid composite revealed significant reduction in wear rate in comparison with Al/AL2O3 composite produced by friction stir processing. Worn surface of the wear test samples were examined by SEM in order to determine wear mechanism.

  2. Effect of Ionizing Radiation on the Mechanical and Structural Properties of Graphite Fiber Reinforced Composites. Ph.D. Thesis

    Science.gov (United States)

    Wolf, Kay Woodroof

    1982-01-01

    Graphite/epoxy (T300/5208) and graphite/polyimide composites (C6000/PMR 15) were exposed to various levels of 0.5 MeV electron radiation with the maximum dose being 10,000 Mrad. A three point bending test was used to evaluate the ultimate stress and modulus of the composites. In all composites except transverse samples of C6000/PMR 15 ultimate stress values remained approximately constant or increased slightly. The modulus values remained approximately constant for all composite types regardless of the radiation level. Interfacial aspects of composites were studied. Interlaminar shear tests were performed on T300/5208 and C6000/PMR 15 composites irradiated to 10,000 Mrad. There was an initial increase in interlaminar shear strength (up to 1,000 Mrad) followed by a sharp decrease with further radiation exposure. Using scanning electron microscopy no visual differences in the mode of fracture could be detected between ruptured control samples and those exposed to various levels of radiation. Electron spectroscopy for chemical analysis (ESCA) revealed little change in the surface elements present in control and highly irradiated T300/5208 composite samples.

  3. Effects of through-the-thickness stitching on impact and interlaminar fracture properties of textile graphite/epoxy laminates

    Science.gov (United States)

    Sharma, Suresh K.; Sankar, Bhavani V.

    1995-01-01

    This study investigated the effects of through-the-thickness stitching on impact damage resistance, impact damage tolerance, and mode I and mode II fracture toughness of textile graphite/epoxy laminates. Uniweave resin-transfer-molded 48 ply graphite/epoxy (AS4/3501-6) laminates were stitched with Kevlar and glass yarns of different linear densities and stitch spacings. Delaminations were implanted during processing to simulate impact damage. Sublaminate buckling tests were performed to determine the effects of stitching on the compressive strength. The results showed outstanding improvements of up to 400 percent in the compression strength over the unstitched laminates. In impact and static indentation tests the onset of damage occurred at the same level, but the extent of damage was less in stitched laminates. Mode I fracture toughness of 24 ply Uniweave unidirectional (AS4/3501-6) stitched laminates was measured by conducting double-cantilever-beam tests. The critical strain energy release rate (G(sub Ic)) was found to be up to 30 times higher than the unstitched laminates. Mode II fracture toughness of the Uniweave laminates was measured by performing end-notched-flexure tests. Two new methods to compute the apparent G(sub IIc) are presented. The apparent G(sub IIc) was found to be at least 5-15 times higher for the stitched laminates.

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

    International Nuclear Information System (INIS)

    Gesh, Christopher J.

    2004-01-01

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

  5. National strategy of nuclear power intellectual property

    International Nuclear Information System (INIS)

    Bo Huaitao

    2008-01-01

    The IP strategy includes four strategic levels: international level, national level, industry level and enterprise level. The national nuclear power IP strategy is an industry level strategy, which is also one part of national IP strategy, permeating with the IP strategy of the international level and enterprise level mutually. Commencing from the angles of the national level and the industry level, the author tries to provide a reference for IP strategy by an initial study about IP strategy in nuclear power industry.. The author holds that independent innovation is at the root of strategic establishment; enterprise is the main body for strategic implementation; and evaluating index must be consummated in the strategic evaluation. (authors)

  6. Magnetic properties of strongly asymmetric nuclear matter

    International Nuclear Information System (INIS)

    Kutschera, M.; Wojcik, W.

    1988-01-01

    We investigate stability of neutron matter containing a small proton admixture with respect to spin fluctuations. We establish conditions under which strongly asymmetric nuclear matter could acquire a permanent magnetization. It is shown that if the protons are localized, the system becomes unstable to spin fluctuations for arbitrarily weak proton-neutron spin interactions. For non-localized protons there exists a threshold value of the spin interaction above which the system can develop a spontaneous polarization. 12 refs., 2 figs. (author)

  7. Influence of irradiation on high-strength graphites

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  8. Porous graphite electrodes for rechargeable ion-transfer batteries

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-06-01

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

  9. Application of neutron activation analysis to the study of impurities in molybdenum, tungsten and nuclear graphite; Application de l'analyse par activation neutronique a l'etude des impuretes dans le molybdene, le tungstene et le graphite nucleaire

    Energy Technology Data Exchange (ETDEWEB)

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

    1967-10-15

    A neutron activation method is described for the analysis of a maximum number of foreign elements in molybdenum, tungsten and graphite. The various elements are isolated using a systematic separation programme; the elements are subsequently analysed qualitatively and quantitatively using {gamma}-spectrometry. By this method are dosed 27 elements in molybdenum and tungsten, and 20 elements in graphite to which can be added those elements which are already the object of routine analysis: V, Mn, Si, P, S, Cl and 14 rare earths. (author) [French] On decrit une methode d'analyse par activation neutronique permettant de doser un maximum d'elements etrangers dans le molybdene, le tungstene et le graphite. En suivant un schema de separation systematique, on isole les differents elements dont les analyses qualitatives et quantitatives sont ensuite effectuees par spectrometrie {gamma}. Par cette methode, on dose 27 elements dans le molybdene et le tungstene, 20 elements dans le graphite, auxquels on peut encore ajouter les elements doses couramment: V, Mn, Si, P, S, Cl et 14 Terres Rares. (auteur)

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

    Directory of Open Access Journals (Sweden)

    Lusitra Munisa

    2003-04-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  12. Preparation and properties of palmitic-stearic acid eutectic mixture/expanded graphite composite as phase change material for energy storage

    International Nuclear Information System (INIS)

    Zhang, Nan; Yuan, Yanping; Du, Yanxia; Cao, Xiaoling; Yuan, Yaguang

    2014-01-01

    A novel composite PCM (phase change material) with PA-SA (palmitic-stearic acid) eutectic mixture as PCM and EG (expanded graphite) as supporting material was prepared. The optimum absorption ratio of PA-SA/EG (Palmitic-stearic acid/expanded graphite) composite PCM was determined as PA-SA:EG = 13:1 (by mass). Scanning electron microscope and Fourier transformation infrared spectroscopy results show that PA-SA was uniformly distributed in the porous network structure of EG due to the physical action. Thermal property and thermal stability of the PA-SA/EG composite PCM were characterized by DSC (differential scanning calorimetry) and TGA (thermogravimetric analysis). DSC results indicated that the melting and freezing temperatures and latent heats of PA-SA/EG were measured as 53.89 °C and 54.37 °C, and 166.27 J/g and 166.13 J/g. TGA test results revealed that PA-SA/EG had a good thermal stability in working temperature range. Thermal cycling test results showed PA-SA/EG had a good thermal reliability after 720 thermal cycles. Thermal conductivity of the composite PCM was measured as 2.51 W/m K, much higher than that of PA-SA. The thermal energy storage and release rates of PA-SA/EG were also increased due to the high thermal conductivity of EG. In conclusion, the prepared PA-SA/EG composite PCM can be acted as a potential material for thermal energy storage due to the acceptable thermal properties, good thermal reliability and stability, high thermal conductivity. - Highlights: • PA-SA/EG (Palmitic-stearic acid/expanded graphite) composite PCM was prepared. • Optimum absorption ratio of PA-SA in EG was obtained as 13:1 (by mass). • Thermal conductivity and performance of PA-SA/EG have been significate improved. • PA-SA/EG has a good thermal reliability and thermal stability

  13. Rapid analysis of 14C and 3H in graphite and concrete for decommissioning of nuclear reactor

    DEFF Research Database (Denmark)

    Hou, Xiaolin

    2005-01-01

    /g graphite and 0.11 and 0.06Bq/g concrete, respectively. The cross contamination of C-14 and tritium in the preparation of samples is less than 0.2%. The interference of other radionuclides in the determination of C-14 and tritium in graphite is insignificant. The analytical accuracy, investigated...

  14. 'In situ' expanded graphite extinguishant

    International Nuclear Information System (INIS)

    Cao Qixin; Shou Yuemei; He Bangrong

    1987-01-01

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

  15. Modeling Fission Product Sorption in Graphite Structures

    International Nuclear Information System (INIS)

    Szlufarska, Izabela; Morgan, Dane; Allen, Todd

    2013-01-01

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

  16. 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

  17. Physical properties of organic nuclear reactor coolants

    Energy Technology Data Exchange (ETDEWEB)

    Elberg, S.; Friz, G.

    1963-03-15

    Diphenyl and terphenyls with different high-boiler content were studied up to temperatures of 450 deg C. Data from high boiler reactors show viscosity (strong influence), thermal conductivity (medium influence), density and specific heat (small influence). The vapor pressure is rn the most affected property (important influence of low boilers). Also viscosity shows an effect. Some data for pure highboilers are also presented. New results were obtained with direct measurements of the latent heat ot vaporization. (P.C.H.)

  18. Properties of form-stable paraffin/silicon dioxide/expanded graphite phase change composites prepared by sol–gel method

    International Nuclear Information System (INIS)

    Li, Min; Wu, Zhishen; Tan, Jinmiao

    2012-01-01

    Highlights: ► Paraffin/SiO 2 /EG composite PCM was prepared with sol–gel method. ► The thermal conductivity of SiO 2 /paraffin/EG is 94.7% higher than paraffin. ► The latent heat of paraffin/SiO 2 /EG composite is 104.4 J/g. -- Abstract: A form-stable paraffin/silicon dioxide (SiO 2 )/expanded graphite (EG) composite phase change material (PCM) was prepared by sol–gel method. Silica gel acts as the supporting material and EG is used to increase the thermal conductivity. The mass fractions of silicon oxide and graphite are 20.8% and 7.2%, respectively. The composite PCM was characterized using scanning electron microscopy (SEM), transmission electron microscopy (TEM) and Fourier transformation infrared spectroscopy (FTIR) method. Thermal properties and thermal stability of the composite PCM were studied using differential scanning calorimetry (DSC). The result shows that paraffin was well dispersed in the network of silica gel and there is no chemical reaction between them. The phase change temperature of the paraffin/SiO 2 composite and the paraffin/SiO 2 /EG composite are 27.53 °C and 27.72 °C, respectively. The latent heat of the paraffin/SiO 2 composite and the paraffin/SiO 2 /EG composite are 112.8 J/g and 104.4 J/g, respectively. The thermal conductivity of the SiO 2 /paraffin composite and the SiO 2 /paraffin/EG composite are 28.2% and 94.7% higher than that of paraffin.

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

    International Nuclear Information System (INIS)

    2000-09-01

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  1. Nuclear properties with realistic Hamiltonians through spectral distribution theory

    International Nuclear Information System (INIS)

    Vary, J.P.; Belehrad, R.; Dalton, B.J.

    1979-01-01

    Motivated by the need of non-perturbative methods for utilizing realistic nuclear Hamiltonians H, the authors use spectral distribution theory, based on calculated moments of H, to obtain specific bulk and valence properties of finite nuclei. The primary emphasis here is to present results for the binding energies of nuclei obtained with and without an assumed core. (Auth.)

  2. Statistical properties of the nuclear shell-model Hamiltonian

    International Nuclear Information System (INIS)

    Dias, H.; Hussein, M.S.; Oliveira, N.A. de

    1986-01-01

    The statistical properties of realistic nuclear shell-model Hamiltonian are investigated in sd-shell nuclei. The probability distribution of the basic-vector amplitude is calculated and compared with the Porter-Thomas distribution. Relevance of the results to the calculation of the giant resonance mixing parameter is pointed out. (Author) [pt

  3. Nuclear wobbling motion and properties of E-2 transitions

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Y R [Kyushu Univ., Fukuoka (Japan). Dept. of Physics; Matsuzaki, M [Fukuoka Univ. (Japan)

    1992-08-01

    The nuclear wobbling motion associated with the static triaxial deformation are discussed based on a microscopic theory. Properties of the E2-transitions between the one-phonon wobbling band and the yrast (vacuum) band are studied and their characteristic features are suggested. (author). 11 refs., 3 figs.

  4. Error estimation for ADS nuclear properties by using nuclear data covariances

    International Nuclear Information System (INIS)

    Tsujimoto, Kazufumi

    2005-01-01

    Error for nuclear properties of accelerator-driven subcritical system by the uncertainties of nuclear data was performed. An uncertainty analysis was done using the sensitivity coefficients based on the generalized perturbation theory and the variance matrix data. For major actinides and structural material, the covariance data in JENDL-3.3 library were used. For MA, newly evaluated covariance data was used since there had been no reliable data in all libraries. (author)

  5. 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

  6. Assisted heterogeneous multinucleation and bubble growth in semicrystalline ethylene-vinyl acetate copolymer/expanded graphite nanocomposite foams: Control of morphology and viscoelastic properties

    Directory of Open Access Journals (Sweden)

    O. Yousefzade

    2015-10-01

    Full Text Available Nanocomposite foams of ethylene-vinyl acetate copolymer (EVA reinforced by expanded graphite (EG were prepared using supercritical nitrogen in batch foaming process. Effects of EG particle size, crosslinking of EVA chains and foaming temperature on the cell morphology and foam viscoelastic properties were investigated. EG sheet surface interestingly provide multiple heterogeneous nucleation sites for bubbles. This role is considerably intensified by incorporating lower loadings of EG with higher aspect ratio. The amorphous and non-crosslinked domains of EVA matrix constitute denser bubble areas. Higher void fraction and more uniform cell structure is achieved for non-crosslinked EVA/EG nanocomposites foamed at higher temperatures. With regard to the structural variation, the void fraction of foam samples decreases with increasing the EG content. Storage and loss moduli were analyzed to study the viscoelastic properties of nanocomposite foams. Surprisingly, the foaming process of EVA results in a drastic reduction in loss and storage moduli regardless of whether the thermoplastic matrix contains EG nanofiller or not. For the EVA/EG foams with the same composition, the nanocomposite having higher void fraction shows relatively lower loss modulus and more restricted molecular movements. The study findings have verified that the dynamics of polymer chains varies after foaming EVA matrix in the presence of EG.

  7. 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

  8. 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)

  9. Nonlinear mean field theory for nuclear matter and surface properties

    International Nuclear Information System (INIS)

    Boguta, J.; Moszkowski, S.A.

    1983-01-01

    Nuclear matter properties are studied in a nonlinear relativistic mean field theory. We determine the parameters of the model from bulk properties of symmetric nuclear matter and a reasonable value of the effective mass. In this work, we stress the nonrelativistic limit of the theory which is essentially equivalent to a Skyrme hamiltonian, and we show that most of the results can be obtained, to a good approximation, analytically. The strength of the required parameters is determined from the binding energy and density of nuclear matter and the effective nucleon mass. For realistic values of the parameters, the nonrelativistic approximation turns out to be quite satisfactory. Using reasonable values of the parameters, we can account for other key properties of nuclei, such as the spin-orbit coupling, surface energy, and diffuseness of the nuclear surface. Also the energy dependence of the nucleon-nucleus optical model is accounted for reasonably well except near the Fermi surface. It is found, in agreement with empirical results, that the Landau parameter F 0 is quite small in normal nuclear matter. Both density dependence and momentum dependence of the NN interaction, but especially the former, are important for nuclear saturation. The required scalar and vector coupling constants agree fairly well with those obtained from analyses of NN scattering phase shifts with one-boson-exchange models. The mean field theory provides a semiquantitative justification for the weak Skyrme interaction in odd states. The strength of the required nonlinear term is roughly consistent with that derived using a new version of the chiral mean field theory in which the vector mass as well as the nucleon mass is generated by the sigma-field. (orig.)

  10. Effect of binder properties on electrochemical performance for silicon-graphite anode: Method and application of binder screening

    International Nuclear Information System (INIS)

    Yim, Taeeun; Choi, Soo Jung; Jo, Yong Nam; Kim, Tae-Hyun; Kim, Ki Jae; Jeong, Goojin; Kim, Young-Jun

    2014-01-01

    Highlights: • Binder properties are systematically characterized to estimate their suitability. • Interpretation of binder properties in connection with binding affinity, electrode properties, and degree of phase separation in slurry. • According to the screening results, hybridization of poly(acrylic acid) and poly(amide imide) is recommended. • The modified binder showed improved cycle performance based on enhanced binder properties. - Abstract: With increasing demand for lithium-ion batteries (LIBs) with high energy density, silicon-based negative electrode material has attracted much interest because of its high specific capacity. Practical utilization of Si remains unattainable, however, owing to severe volume expansion in the electrode, resulting in a loss of the electrical Si network, which is directly connected to drastic capacity fading of the cell. Therefore, there have been systematic studies on the characterization of fundamental binder properties to estimate the suitability of various binder materials. The binder properties are subdivided into mechanical and adhesion characteristics, electrode properties (rigidity and recovery), and phase separation behavior of slurry to correlate with the electrochemical performance and practical acceptance of candidate materials. Systematic screening showed that hybridization of poly(acrylic acid) (PAA) and poly(amide imide) (PAI) could complement each other's properties and the hybridized PAA–PAI was synthesized by a one-step, acid-catalyzed reaction. The PAA–PAI hybrid showed enhancement in overall properties as a result of co-polymerization and exhibited remarkable cycling performance after 300 cycles. Based on these results, it can be concluded that an understanding of binder characteristics provides useful insight into the search for a more efficient binder material, and fine tuning of fundamental binder properties through screening will be advantageous to the construction of more efficient LIB

  11. Damage, trapping and desorption at the implantation of helium and deuterium in graphite, diamond and silicon carbide

    International Nuclear Information System (INIS)

    Lopez, G.A.R.

    1995-07-01

    The production, thermal stability and structure of ion induced defects have been studied by Rutherford backscattering in channeling geometry for the implantation of helium and deuterium in graphite, diamond and silicon carbide with energies of 8 and 20 keV. At the implantation of deuterium and helium ions more defects were measured in graphite than in diamond or silicon carbide at equal experimental conditions. This is due to increased backscattering in graphite, which is caused by the splitting and tilting of crystallites and a local reordering of lattice atoms around defects. At 300 K, Helium produces more defects in all three materials than deuterium with equal depth distribution of defects. The ratio of the defects produced by helium and deuterium agrees very well with the corresponding ratio of the energy deposited in nuclear collisions. In graphite, only small concentrations of deuterium induced defects anneal below 800 K, while in diamond small concentrations of deuterium as well as of helium induced defects anneal mostly below 800 K. This annealing behavior is considered to be due to recombination of point defects. The buildup of helium and deuterium in graphite is different. The trapping of deuterium proceeds until saturation is reached, while in the case of helium trapping is interrupted by flaking. In diamond, deuterium as well as helium are trapped almost completely until at higher fluences reemission starts and saturation is reached. Two desorption mechanisms were identified for the thermal desorption of helium from base-oriented graphite. Helium implanted at low fluences desorbs diffusing to the surface, while for the implantation of high fluences the release of helium due to blistering dominates. The desorption of deuterium from graphite and diamond shows differences. While in graphite the desorption starts already at 800 K, in diamond up to 1140 K only little desorption can be observed. These differences can be explained by the different transport

  12. Probing properties of neutron stars with terrestrial nuclear reactions

    International Nuclear Information System (INIS)

    Li Baoan; Chen Liewen; Ko, C.M.; Steiner, Andrew W.; Yong Gaochan

    2006-01-01

    Heavy-ion reactions induced by neutron-rich nuclei provide the unique opportunity in terrestrial laboratories to constrain the nuclear symmetry energy Esym in a broad density range. A conservative constraint, 32(ρ/ρ0)0.7 < Esym(ρ) < 32(ρ/ρ0)1.1, around the nuclear matter saturation density ρ0 has recently been obtained from analyzing the isospin diffusion data within a transport model for intermediate energy heavy-ion reactions. This subsequently puts a stringent constraint on properties of neutron stars, especially their radii and cooling mechanisms

  13. Thermal properties of nuclear matter under the periodic boundary condition

    International Nuclear Information System (INIS)

    Otuka, Naohiko; Ohnishi, Akira

    1999-01-01

    We present the thermal properties of nuclear matter under the periodic boundary condition by the use of our hadronic nucleus-nucleus cascade model (HANDEL) which is developed to treat relativistic heavy-ion collisions from BNL-AGS to CERN-SPS. We first show some results of p-p scattering calculation in our new version which is improved in order to treat isospin ratio and multiplicity more accurately. We then display the results of calculation of nuclear matter with baryon density ρ b = 0.77 fm 3 at some energy densities. Time evolution of particle abundance and temperature are shown. (author)

  14. Nuclear power plants after Angra-3. Property and operation flexibility

    International Nuclear Information System (INIS)

    Deppe, Lothario

    2008-01-01

    The present work analyses the consequences that will be introduced with the modifications in the actual existing systematic related to the property and operation of nuclear power plants, coming from the legislation of 2004 that establishes that the electric energy distribution concessionaires acquire energy for attending its markets through bid in the modality of auction conducted by ANEEL - Brazilian Agency of Electric Energy. This systematic means that probably the Nuclear Power Plants after Angra-3 will be submitted to this auction process. The present report analyses all these questions and proposes solutions that seems to be the most obvious and viable

  15. Statistical nuclear properties and synthesis of 138La

    Directory of Open Access Journals (Sweden)

    Kheswa B. V.

    2015-01-01

    Full Text Available The synthesis of the neutron deficient 138La nucleus has been a puzzle for a long time. It has not been clear whether it is produced through photodisintegration processes or neutrino induced reactions due to unavailability of experimental data for nuclear level densities and γ strength functions of 138,139La nuclei. In the present work these nuclear properties have been measured and are used to investigate the synthesis of 138La. The results support the neutrino interactions as a dominant production process for 138La.

  16. 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)

  17. Effect of gamma radiation on graphite - PTFE dry lubrication system

    Science.gov (United States)

    Singh, Sachin; Tyagi, Mukti; Seshadri, Geetha; Tyagi, Ajay Kumar; Varshney, Lalit

    2017-12-01

    An effect of gamma radiation on lubrication behavior of graphite -PTFE dry lubrication system has been studied using (TR-TW-30L) tribometer with thrust washer attachment in plane contact. Different compositions of graphite and PTFE were prepared and irradiated by gamma rays. Gamma radiation exposure significantly improves the tribological properties indicated by decrease in coefficient of friction and wear properties of graphite -PTFE dry lubrication system. SEM and XRD analysis confirm the physico-chemical modification of graphite-PTFE on gamma radiation exposure leading to a novel dry lubrication system with good slip and anti friction properties.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-08-14

    Superhydrophilic graphite surfaces and water-dispersible graphite colloids are obtained by electrochemical exfoliation with hydrophobic graphite electrodes. Such counterintuitive characteristics are caused by partial oxidation and investigated by examining both graphite electrodes and exfoliated particles after electrolysis. The extent of surface oxidation can be explored through contact angle measurement, scanning electron microscope, electrical sheet resistance, x-ray photoelectron spectroscopy, zeta-potential analyzer, thermogravimetric analysis, UV-visible, and Raman spectroscopy. The degree of wettability of the graphite anode can be altered by the electrolytic current and time. The water contact angle declines generally with increasing the electrolytic current or time. After a sufficient time, the graphite anode becomes superhydrophilic and its hydrophobicity can be recovered by peeling with adhesive tape. This consequence reveals that the anodic graphite is oxidized by oxygen bubbles but the oxidation just occurs at the outer layers of the graphite sheet. Moreover, the characteristics of oxidation revealed by UV peak shift, peak ratio between D and G bands, and negative zeta-potential indicate the presence of graphite oxide on the outer shell of the exfoliated colloids. However, thermogravimetric analysis for the extent of decomposition of oxygen functional groups verifies that the amount of oxygen groups is significantly less than that of graphite oxide prepared via Hummer method. The structure of this partially oxidized graphite may consist of a graphite core covered with an oxidized shell. The properties of the exfoliated colloids are also influenced by pH of the electrolytic solution. As pH is increased, the extent of oxidation descends and the thickness of oxidized shell decreases. Those results reveal that the degree of oxidation of exfoliated nanoparticles can be manipulated simply by controlling pH.

  19. 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

  20. Thermostatic properties of semi-infinite polarized nuclear matter

    International Nuclear Information System (INIS)

    Abd-Alla, M.; Hassan, M.Y.M.; Ramadan, S.

    1988-03-01

    The surface and curvature properties of semi-infinite polarized nuclear matter (SPNM) are calculated using an expansion for the Fermi integrals up to T 2 . A density matrix expansion is obtained for a modified form of Seyler-Blanchard interaction. New parameters that characterize the surface and curvature properties of SPNM are introduced. The level density parameter is extracted from the low temperature expansion of the free energy and compared with previous calculations. A reasonable agreement is obtained for the parameters calculated before. (author). 78 refs, 1 fig., 5 tabs

  1. A Study Of Physical Properties Of Matrix Graphite Particle's Distribution As Ras Materials On 900oC Baking Stage Temperature

    International Nuclear Information System (INIS)

    Sajekti, Kasilani Noor; Dahroni, Imam; Nawangsih, Endang

    2000-01-01

    To aim's the physical characteristic of matrix graphite materials the physical basic characteristics were necessary prepared. Investigation of calsine cokes particle size distribution effect on 900 o C temperature baking stage had been done. The calsine coke and tar pitch were crushed and sieved, to get a particle size 63; 90; 106 and 125 μm, making pellet by mixed with 33% weight of tar pitch, than grilled at 900 o C during 30 minutes. Grilled products: physical (bulk density, electrical resistivity), mechanics (hardness, compressive strength) and micro's (surface area, total and pore radius) were analyzed. From the 9 samples, 3 samples in fulfilled condition with mixed particle size of calsine cokes 63 and 106 μm and the best weight ratio between calsine coke and tar pitch were 2/3:4/3 to 1. The physical properties yield were 1.19 g/mL bulk density, the electrical resistivity 2.63 Ωcm, the hardness 5.90 kg/mm 2 , the compressive strength 1600 Newton, the density (N 2 adsorbate) 2.89 g/mL, the specific surface area 8.08 mm 2 /g,the total pore /volume 1.48% and the average pore radius 12.60 Angstrom

  2. Preparation and electrochemical property of TiO_2/Nano-graphite composite anode for electro-catalytic degradation of ceftriaxone sodium

    International Nuclear Information System (INIS)

    Guo, Xiaolei; Li, Dong; Wan, Jiafeng; Yu, Xiujuan

    2015-01-01

    Titanium dioxide/Nano-graphite (TiO_2/Nano-G) composite was synthesized by a sol-gel method and TiO_2/Nano-G electrode was prepared in hot-press approach. The composite was characterized by X-ray photoelectron spectroscopy (XPS), fourier transform infrared (FT-IR), scanning electrons microscopy (SEM), transmission electron microscopy (TEM) and X-ray diffraction (XRD). The electrochemical performance of the TiO_2/Nano-G anode electrode was investigated through cyclic voltammetry (CV) and electrochemical impedance spectroscopy (EIS). The electro-catalytic performance was evaluated by the yield of ·OH radicals, degradation of methyl orange and ceftriaxone sodium. The results demonstrated that TiO_2 nanoparticles were dispersed on the surface and interlamination of the Nano-G uniformly, TiO_2/Nano-G electrode owned higher electro-catalytic oxidation activity and stability than Nano-G electrode. Degradation rate of ceftriaxone sodium within 120 min by TiO_2(40)/Nano-G electrode was 97.7%. And ·OH radicals given by TiO_2/Nano-G electrode was higher than that of Nano-G electrode and DSA (Ti/IrO_2-RuO_2) electrode. The excellent electro-catalytic performance could be ascribed to the admirable conductive property of the Nano-G and more production of ·OH offered by TiO_2(40)/Nano-G electrode.

  3. Thermodynamic and structural properties of ball-milled mixtures composed of nano-structural graphite and alkali(-earth) metal hydride

    International Nuclear Information System (INIS)

    Miyaoka, Hiroki; Ichikawa, Takayuki; Fujii, Hironobu

    2007-01-01

    Hydrogen desorption properties of mechanically milled materials composed of nano-structural hydrogenated-graphite (C nano H x ) and alkali(-earth) metal hydride (MH; M = Na, Mg and Ca) were investigated from the thermodynamic and structural points of view. The hydrogen desorption temperature for all the C nano H x and MH composites was obviously lower than that of the corresponding each hydride. In addition, the desorption of hydrocarbons from C nano H x was significantly suppressed by making composite of C nano H x with MH, even though C nano H x itself thermally desorbs a considerably large amount of hydrocarbons. These results indicate that an interaction exists between C nano H x and MH, and hydrogen in both the phases is destabilized by a close contact between polar C-H groups in C nano H x and the MH solid phase. Moreover, a new type of chemical bonding between the nano-structural carbon (C nano ) and the Li, Ca, or Mg metal atoms may be formed after hydrogen desorption. Thus, the above metal-C-H system would be recognized as a new family of H-storage materials

  4. Effect of boron and phosphorus codoping on the electronic and optical properties of graphitic carbon nitride monolayers: First-principle simulations

    Science.gov (United States)

    Yousefi, Mahdieh; Faraji, Monireh; Asgari, Reza; Moshfegh, Alireza Z.

    2018-05-01

    We study the effect of boron (B) and phosphorous (P) doping and B/P codoping on electronic and optical properties of graphitic carbon nitride (g-C3N4 or GCN) monolayers using density functional simulations. The energy band structure indicates that the incorporation of both B and P into a hexagonal lattice of GCN reduces the energy band gap from 3.1 for pristine GCN to 1.9 eV, thus extending light absorption toward the visible region. Moreover, on the basis of calculating absorption spectra and dielectric function, the codoped system exhibits an improved absorption intensity in the visible region and more electronic transitions, which named π* electronic transitions that occurred and were prohibited in the pristine GCN. These transitions can be attributed to charge redistribution upon doping, caused by distorted configurable B/P-codoped GCN confirmed by both electron density and Mulliken charge population. Therefore, B/P-codoped GCN is expected to be an auspicious candidate to be used as a promising photoelectrode in photoelectrochemical water splitting reactions leading to efficient solar H2 production.

  5. The influence of expanded graphite on thermal properties for paraffin/high density polyethylene/chlorinated paraffin/antimony trioxide as a flame retardant phase change material

    International Nuclear Information System (INIS)

    Zhang Ping; Song Lei; Lu Hongdian; Wang Jian; Hu Yuan

    2010-01-01

    The influences of expanded graphite (EG) on the thermal properties of chlorinated paraffin (CP) and antimony trioxide (AT) on phase change material which bases on paraffin/high density polyethylene (HDPE) are studied. Differential scanning calorimetry (DSC), thermogravimetric analysis (TGA), thermogravimetric analysis-Fourier transform infrared spectrometry (TGA-FTIR), microscale combustion calorimeter (MCC) and cone calorimeter (CONE) were used to evaluate the influence of EG on paraffin/HDPE/CP/AT system. The DSC results indicated that the latent heat value of PCM could be increased when the mass fraction of HDPE was decreased in the PCM, and EG could confine the molecular heat movement of paraffin. EG could improve the thermal stability and increase the char residue at high temperature for paraffin/HDPE/CP/AT hybrid. The volatilized products formed on thermal degradation of paraffin/HDPE/CP/AT with EG showed the release of CO 2 gas was hastened and increased, and the amount of combustible gases were decreased by TGA-FTIR analysis. The MCC and CONE results presented that the flame retardant efficiency of CP/AT could be improved by adding EG in paraffin/HDPE/CP/AT system.

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

    International Nuclear Information System (INIS)

    Hoinkis, E.

    1991-03-01

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

  7. 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)

  8. Graphite moderated 252Cf source

    International Nuclear Information System (INIS)

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

    2014-08-01

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

  9. High temperature tests for graphite materials

    OpenAIRE

    Zhmurikov, Evgenij

    2015-01-01

    This study was performed within the framework of the EURISOL for facilities SPIRAL-II (GANIL, France) and SPES (LNL, Italy), and aims to investigate the anticipated strength properties of fine-grained graphite at elevated temperatures. It appears that the major parameters that affect to the lifetime of a graphite target of this IP are the temperature and heating time. High temperature tests were conducted to simulate the heating under the influence of a beam of heavy particles by passing thro...

  10. Actinides in irradiated graphite of RBMK-1500 reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-01

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

  11. 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

  12. 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)

  13. Temperature and irradiation effects on the behaviour of 14C and its precursor 14N in nuclear graphite. Study of a decontamination process using steam reforming

    International Nuclear Information System (INIS)

    Silbermann, Gwennaelle

    2013-01-01

    The dismantling of UNGG reactors in France will generate about 23 000 tons of radioactive graphite wastes. To manage these wastes, the radiological inventory and data on radionuclides (RN) location and speciation should be determined. 14 C was identified as an important RN for disposal due to its high initial activity and the risk of release of a mobile organic fraction in environment, after water ingress into the disposal. Hence, the objective of this thesis, carried out in partnership with EDF is to implement experimental studies to simulate and evaluate the impact of temperature, irradiation and graphite radiolytic corrosion on the in reactor behavior of 14 C and its precursor, 14 N. The obtained data are then used to study the thermal decontamination of graphite in presence of water vapor. The experimental approach aims at simulating the presence of 14 C and 14 N by the respective ion implantation of 13 C and 14 N or 15 N in virgin graphite. This study shows that, in the temperature range reached during reactor operation, (100-500 C) and without radiolytic corrosion, 13 C is thermally stable whatever the initial graphite structure. Moreover, irradiation experiments were performed on heated graphite (500 C) put in contact with a gas representative of the radiolized coolant gas. They show the synergistic role played by the oxidative species and the graphite structure disorder on the enhancement of 13 C mobility resulting in the gasification of the graphite surface and/or the selective oxidation of 13 C more weakly bound than 12 C. Concerning the pristine nitrogen, we showed first that the surface concentration reaches several hundred ppm (≤500 ppm at) and decreases at deeper depths to about 160 ppm at.. Unlike implanted 13 C, implanted nitrogen migrates at 500 C when the graphite is highly disordered (about 8 dpa) while remaining stable for a lower disorder rate (0.14 dpa). Experiments also show the synergistic role by electronic excitations and temperature

  14. A graphite nanoeraser

    DEFF Research Database (Denmark)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  16. Investigation on Electrochemical Properties of Polythiophene Nanocomposite with Graphite Derivatives as Supercapacitor Material on Breath Figure-Decorated PMMA Electrode

    Science.gov (United States)

    Azimi, Mona; Abbaspour, Mohsen; Fazli, Ali; Setoodeh, Hamideh; Pourabbas, Behzad

    2018-03-01

    Breath figures have been formed by the direct breath figure method on polymethyl methacrylate electrode sand hexagonal oriented holes with 0.5- to 10- μm2 surface area have been created. Deposition of materials on the electrodes has been performed by the spray-coating method. polythiophene (PTh) nanoparticles, polythiophene-graphene oxide (PTh-GO) and polythiophene-reduced graphene oxide (PTh-G) nanocomposites were synthesized by emulsion polymerization, while characterization of synthetic materials have been carried out by Fourier transform infrared, Χ-ray diffraction, transmission electron microscopy, UV-Vis spectroscopy and field emission scanning electron microscopy techniques. Also, the electrochemical properties of the designed electrodes were investigated by cyclic voltammetry, galvanostatic charge-discharge and electrochemical impedance spectroscopy techniques. Specific capacitance of porous electrodes coated by PTh nanoparticles, PTh-GO and PTh-G nanocomposites were calculated from cyclic voltammetry curves at 5 mV/s scan rate, andthe values are 3.5 F/g, 16.39 F/g, and 28.68 F/g, respectively. Also, the energy density of each electrode at 5 mV/s scan rate has been calculated and the results show that incorporation of GO and G nanolayers with PTh nanoparticles enhances the electrochemical properties of electrodes.

  17. 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

  18. 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

  19. Fabrication of conducting composite sheets using cost-effective graphite flakes and amorphous styrene acrylonitrile for enhanced thermistor, dielectric, and electromagnetic interference shielding properties

    Energy Technology Data Exchange (ETDEWEB)

    Panwar, Varij, E-mail: varijpanwarcertain@gmail.com [Electronics and Communication Engineering, Graphic Era University, Dehradun, Uttarakhand (India); Gill, Fateh Singh; Rathi, Vikas; Tewari, V.K. [Electronics and Communication Engineering, Graphic Era University, Dehradun, Uttarakhand (India); Mehra, R.M. [Sharda University, Greater Noida (India); Park, Jong-Oh, E-mail: jop@jnu.ac.kr [School of Mechanical Engineering, Chonnam National University, Gwangju (Korea, Republic of); Park, Sukho, E-mail: shpark12@dgist.ac.kr [Department of Robotics Engineering, Daegu Gyeongbuk Institute of Science and Technology, Daegu (Korea, Republic of)

    2017-06-01

    The fabrication of strong conducting composite sheets (CCSs) using a simple technique with cost-effective materials is desirable for capacitor, decoupling capacitor, and electromagnetic interference (EMI) shielding applications. Here, we used cost-effective graphite flakes (GFs) as a conducting filler and amorphous poly (styrene-co-acrylonitrile) (PSAN) as an insulating polymer to fabricate a CCS via a simple mechanical mixing and hot compression molding process in 2.5 h, with the aim to save time and avoid the use of toxic reagents, which are generally used in chemical methods. In the present method, the GFs are connected in diffusively adhere polymer matrix, controlled by temperature and pressure that generate the conduction in the CCSs. The resulting PSAN/GF CCSs were characterized by using scanning electron microscopy (SEM), differential scanning calorimetry (DSC), thermogravimetric analysis (TGA), and hardness tests. The GFs penetrated the interfacial region of PSAN, thus improving the thermistor and dielectric properties (dielectric constant, AC conductivity, and dissipation factor) of the PSAN/GF CCSs. Furthermore, the PSAN/GF CCSs showed enhanced hardness and EMI shielding effectiveness (SE) properties in the X-band frequency range (8.5–12.5 GHz). The percolation theory was implemented to DC and AC conductivity. To detect the transition of the dielectric properties, the dielectric constant of the CCSs was analyzed with increasing volume fraction of GFs in the radio frequency region. The improved dielectric constant, AC conductivity, and dissipation factor of the PSAN/GF CCS, indicated a significant improvement in their EMI shielding properties in the X-band frequency range, which were measured using the waveguide method. The ac conductivity of PSAN/GF CCS shows stable behavior in the higher frequency ranges. The EMISE of PSAN/GF CCS were found to increase with increasing GF content due to the absorbance mechanism. - Highlights: • Enhanced hardness and

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

    Directory of Open Access Journals (Sweden)

    Bespala Evgeny

    2016-01-01

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

  1. Study on practical of eddy current testing of core and core support graphite components in HTTR

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Iyoku, Tatsuo; Ooka, Norikazu; Shindo, Yoshihisa; Kawae, Hidetoshi; Hayashi, Motomitsu; Kambe, Mamoru; Takahashi, Masaaki; Ide, Akira.

    1994-01-01

    Core and core support graphite components in the HTTR (High Temperature Engineering Test Reactor) are mainly made of nuclear-grade IG-110 and PGX graphites. Nondestructive inspection with Eddy Current Testing (ECT) is planned to be applied to these components. The method of ECT has been already established for metallic components, however, cannot be applied directly to the graphite ones, because the characteristics of graphite are quite different in micro-structure from those of metals. Therefore, ECT method and condition were studied for the application of the ECT to the graphite components. This paper describes the study on practical method and conditions of ECT for above mentioned graphite structures. (author)

  2. Self-consistent determination of quasiparticle properties in nuclear matter

    International Nuclear Information System (INIS)

    Oset, E.; Palanques-Mestre, A.

    1981-01-01

    The self-energy of nuclear matter is calculated by directing the attention to the energy and momentum dependent pieces which determine the quasiparticle properties. A microscopic approach is followed which starts from the boson exchange picture for the NN interaction, then the π-and p-mesons are shown to play a major role in the nucleon renormalization. The calculation is done self-consistently and the effective mass and pole strength determined as a function of the nuclear density and momentum. Particular emphasis is put on the non-static character of the interaction and its consequences. Finally a comparison is made with other calculations and with experimental results. The consequences of the nucleon renormalization in pion condensation are also examined with the result that the critical density is pushed up appreciably. (orig.)

  3. Quantitative properties of clustering within modern microscopic nuclear models

    International Nuclear Information System (INIS)

    Volya, A.; Tchuvil’sky, Yu. M.

    2016-01-01

    A method for studying cluster spectroscopic properties of nuclear fragmentation, such as spectroscopic amplitudes, cluster form factors, and spectroscopic factors, is developed on the basis of modern precision nuclear models that take into account the mixing of large-scale shell-model configurations. Alpha-cluster channels are considered as an example. A mathematical proof of the need for taking into account the channel-wave-function renormalization generated by exchange terms of the antisymmetrization operator (Fliessbach effect) is given. Examples where this effect is confirmed by a high quality of the description of experimental data are presented. By and large, the method in question extends substantially the possibilities for studying clustering phenomena in nuclei and for improving the quality of their description.

  4. Development of integrated waste management options for irradiated graphite

    Directory of Open Access Journals (Sweden)

    Alan Wareing

    2017-08-01

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

  5. Preparation of in-house graphite reference material for boron

    International Nuclear Information System (INIS)

    Kumar, Sanjukta A.; Venkatesh, K.; Swain, Kallola K.; Manisha, V.; Kamble, Granthali S.; Pandey, Shailaja P.; Remya Devi, P.S.; Ghosh, M.; Verma, R.

    2016-05-01

    Graphite is extensively used in nuclear technology. Boron concentration in graphite is one of the important parameters that decide its acceptability for nuclear applications. Reliable analytical methods are essential for the determination of boron in graphite at concentration about 5 mg kg -1 . Reference materials are used for validation of existing analytical methods and developing new methodologies. In view of the importance of determination of boron in graphite and unavailability of graphite reference material, an In-house graphite reference material was prepared in Analytical Chemistry Division. Graphite source material was procured, processed to obtain powder of ≤ 75 μm (200 mesh) and bottled. Procedures were developed for the determination of boron in graphite using inductively coupled plasma optical emission spectrometry (ICPOES) and inductively coupled plasma mass spectrometry (ICPMS) techniques. Homogeneity testing was carried out on the bottled units and boron content along with the combined and expanded uncertainties were established. The assigned boron concentration in the In-house graphite reference material is (7.3±0.46) mg kg -1 . (author)

  6. Development of integrated waste management options for irradiated graphite

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-15

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

  7. Property-process relationships in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Tikare, V.

    2015-01-01

    Nuclear fuels are fabricated using many different techniques as they come in a large variety of shapes and compositions. The design and composition of nuclear fuels are predominantly dictated by the engineering requirements necessary for their function in reactors of various designs. Other engineering properties requirements originate from safety and security concerns, and the easy of handling, storing, transporting and disposing of the radioactive materials. In this chapter, the more common of these fuels will be briefly reviewed and the methods used to fabricate them will be presented. The fuels considered in this paper are oxide fuels used in LWRs and FRs, metal fuels in FRs and particulate fuels used in HTGRs. Fabrication of alternative fuel forms and use of standard fuels in alternative reactors will be discussed briefly. The primary motivation to advance fuel fabrication is to improve performance, reduce cost, reduce waste or enhance safety and security of the fuels. To achieve optimal performance, developing models to advance fuel fabrication has to be done in concert with developing fuel performance models. The specific properties and microstructures necessary for improved fuel performance must be identified using fuel performance models, while fuel fabrication models that can determine processing variables to give the desired microstructure and materials properties must be developed. (author)

  8. The influence of target properties on nuclear spectroscopy measurements

    International Nuclear Information System (INIS)

    Dionisio, J.S.; Vieu, C.; Lagrange, J.M.; Pautrat, M.; Vanhorenbeeck, J.; Passoja, A.

    1988-01-01

    A broad review of different kinds of in-beam nuclear spectroscopy measurements particularly influenced by the target properties is outlined. To illustrate such an influence a few typical examples of in-beam electron and gamma-ray spectroscopy measurements, performed at the Orsay MP Tandem accelerator, are reported. In particular several applications of the recoil ion catcher method in the study of short-lived nuclear isomers (with half-lives between ten and few hundred nanoseconds) are briefly described. This method is operated mostly with a pulsed heavy ion beam, bombarding a thin self-supported target but avoiding hitting the catcher foil. Moreover, the time of flight filtering properties of this experimental device is improved by a fast detection of compound nucleus deexcitation (performed with an array of several BaF 2 crystals). This kind of measurement shows clearly the importance of the target qualities as well as the need of good focusing properties and time structure for the accelerated particle beam. Finally, the required characteristics of the targets and recoil stopper foils needed for these measurements (and similar ones performed with the recoil ion shadow method) are analyzed in detail for a few typical experimental arrangements. (author). Abstract only

  9. Monovacancy paramagnetism in neutron-irradiated graphite probed by 13C NMR.

    Science.gov (United States)

    Zhang, Z T; Xu, C; Dmytriieva, D; Molatta, S; Wosnitza, J; Wang, Y T; Helm, M; Zhou, Shengqiang; Kühne, H

    2017-10-20

    We report on the magnetic properties of monovacancy defects in neutron-irradiated graphite, probed by 13 C nuclear magnetic resonance spectroscopy. The bulk paramagnetism of the defect moments is revealed by the temperature dependence of the NMR frequency shift and spectral linewidth, both of which follow a Curie behavior, in agreement with measurements of the macroscopic magnetization. Compared to pristine graphite, the fluctuating hyperfine fields generated by the defect moments lead to an enhancement of the 13 C nuclear spin-lattice relaxation rate [Formula: see text] by about two orders of magnitude. With an applied magnetic field of 7.1 T, the temperature dependence of [Formula: see text] below about 10 K can well be described by a thermally activated form, [Formula: see text], yielding a singular Zeeman energy of ([Formula: see text]) meV, in excellent agreement with the sole presence of polarized, non-interacting defect moments.

  10. Optical motion control of maglev