WorldWideScience

Sample records for ig-110 graphite irradiation

  1. Neutron irradiation induced microstructural changes in NBG-18 and IG-110 nuclear graphites

    Energy Technology Data Exchange (ETDEWEB)

    Karthik, Chinnathambi [Boise State Univ., ID (United States). Dept. of Materials Science and Engineering; Idaho Falls, ID (United States) Center for Advanced Energy Studies; Kane, Joshua [Boise State Univ., ID (United States). Dept. of Materials Science and Engineering; Idaho Falls, ID (United States) Center for Advanced Energy Studies; Idaho National Lab. (INL), Idaho Falls, ID (United States); Butt, Darryl P. [Boise State Univ., ID (United States). Dept. of Materials Science and Engineering; Idaho Falls, ID (United States) Center for Advanced Energy Studies; Windes, William E. [Idaho Falls, ID (United States) Center for Advanced Energy Studies; Idaho National Lab. (INL), Idaho Falls, ID (United States); Ubic, Rick [Boise State Univ., ID (United States). Dept. of Materials Science and Engineering; Idaho Falls, ID (United States) Center for Advanced Energy Studies

    2015-05-01

    This paper reports the neutron-irradiation-induced effects on the microstructure of NBG-18 and IG-110 nuclear graphites. The high-temperature neutron irradiation at two different irradiation conditions was carried out at the Advanced Test Reactor National User Facility at the Idaho National Laboratory. NBG-18 samples were irradiated to 1.54 dpa and 6.78 dpa at 430 °C and 678 °C respectively. IG-110 samples were irradiated to 1.91 dpa and 6.70 dpa at 451 °C and 674 °C respectively. Bright-field transmission electron microscopy imaging was used to study the changes in different microstructural components such as filler particles, microcracks, binder and quinoline-insoluble (QI) particles. Significant changes have been observed in samples irradiated to about 6.7 dpa. The closing of pre-existing microcracks was observed in both the filler and the binder phases. The binder phase exhibited substantial densification with near complete elimination of the microcracks. The QI particles embedded in the binder phase exhibited a complete microstructural transformation from rosettes to highly crystalline solid spheres. The lattice images indicate the formation of edge dislocations as well as extended line defects bridging the adjacent basal planes. The positive climb of these dislocations has been identified as the main contributor to the irradiation-induced swelling of the graphite lattice.

  2. Characterization of structural defects in nuclear graphite IG-110 and NBG-18

    International Nuclear Information System (INIS)

    Zheng, Guiqiu; Xu, Peng; Sridharan, Kumar; Allen, Todd

    2014-01-01

    Nuclear graphite IG-110 and NBG-18 were examined using X-ray diffraction (XRD), Raman spectroscopy, scanning electron microscope (SEM) and high resolution transmission electron microscope (HR-TEM) to understand the structure and microstructure of nuclear graphite. The lattice parameter (a), degree of graphitization (g ¯ ), crystallite size parallel and perpendicular to c-direction (L c and L ⊥ ), anisotropy (B), as well as in-plane crystallite size (L a ) were calculated and compared based on XRD patterns and Raman spectra. Results indicate that IG-110 has a larger crystallite size and higher degree of graphitization, but lower anisotropy than NBG-18. These differences are attributed to the properties of coke source and manufacturing processes. Additionally, the shape of the pores and crystallized filler particles, the interface between binders and fillers, Mrozowski cracks and nano-cracks, and the defects of disclination were observed and characterized from SEM and HR-TEM images. The similarities and differences in microstructure between IG-110 and NBG-18 are discussed. The results in this work provide useful information to guide selection of nuclear graphite for the design of next generation nuclear plants (NGNP)

  3. ICP-MS measurement of silver diffusion coefficient in graphite IG-110 between 1048K and 1284K

    Science.gov (United States)

    Carter, L. M.; Seelig, J. D.; Brockman, J. D.; Robertson, J. D.; Loyalka, S. K.

    2018-01-01

    Silver-110m has been shown to permeate intact silicon carbide and pyrolytic carbon coating layers of the TRISO fuel particles during normal High Temperature Gas-Cooled Reactor (HTGR) operational conditions. The diffusion coefficients for silver in graphite IG-110 measured using a release method designed to simulate HTGR conditions of high temperature and flowing helium in the temperature range 1048-1253 K are reported. The measurements were made using spheres milled from IG-110 graphite that were infused with silver using a pressure vessel technique. The Ag diffusion was measured using a time release technique with an ICP-MS instrument for detection. The results of this work are:

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  5. Irradiation creep of graphite

    International Nuclear Information System (INIS)

    Kennedy, C.R.

    1990-01-01

    Displacement damage of graphite by neutron irradiation causes graphite to change dimensions. This dimensional instability requires careful attention when graphite is used as as moderator and reflector material in nuclear devices. Natural gradients in flux and temperature result in time-varying differential growth generating stresses similar to thermal stresses with an ever increasing temperature gradient. Graphite, however, does have the ability to creep under irradiation, allowing the stress intensity to relax below the fracture strength of the material. Creep strain also serves to average the radiation-induced strains, thus contributing to the stability of the core. As the dimensional instability is a function of temperature, so are the creep characteristics of graphite, and it is of interest to generalize the available data for extension to more extreme conditions of fluence and temperature. Irradiation creep of graphite is characterized by two stages of creep; a primary stage that saturates with time and a secondary stage that is generally assumed to be linear and constant with time. Virtually all past studies have not considered primary creep in detail primarily because there is limited available data at the very low fluences required to saturate primary creep. It is the purpose of this study to carefully examine primary creep in detail over the irradiation temperature range of 150 to 1000 degree C. These studies also include the combined effects of creep, differential growth, and structural changes in graphite by irradiation. 3 refs., 5 figs

  6. 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.

  7. (Irradiation creep of graphite)

    Energy Technology Data Exchange (ETDEWEB)

    Kennedy, C.R.

    1990-12-21

    The traveler attended the Conference, International Symposium on Carbon, to present an invited paper, Irradiation Creep of Graphite,'' and chair one of the technical sessions. There were many papers of particular interest to ORNL and HTGR technology presented by the Japanese since they do not have a particular technology embargo and are quite open in describing their work and results. In particular, a paper describing the failure of Minor's law to predict the fatigue life of graphite was presented. Although the conference had an international flavor, it was dominated by the Japanese. This was primarily a result of geography; however, the work presented by the Japanese illustrated an internal program that is very comprehensive. This conference, a result of this program, was better than all other carbon conferences attended by the traveler. This conference emphasizes the need for US participation in international conferences in order to stay abreast of the rapidly expanding HTGR and graphite technology throughout the world. The United States is no longer a leader in some emerging technologies. The traveler was surprised by the Japanese position in their HTGR development. Their reactor is licensed and the major problem in their graphite program is how to eliminate it with the least perturbation now that most of the work has been done.

  8. 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

  9. 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

  10. 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...

  11. Comparison of NBG-18, NBG-17, IG-110 and IG-11 oxidation kinetics in air

    Science.gov (United States)

    Lee, Jo Jo; Ghosh, Tushar K.; Loyalka, Sudarshan K.

    2018-03-01

    The oxidation rates of several nuclear-grade graphites, NBG-18, NBG-17, IG-110 and IG-11, were measured in air using thermogravimetry. Kinetic parameters and oxidation behavior for each grade were compared by coke type, filler grain size and microstructure. The thickness of the oxidized layer for each grade was determined by layer peeling and direct density measurements. The results for NBG-17 and IG-11 were compared with those available in the literature and our recently reported results for NBG-18 and IG-110 oxidation in air. The finer-grained graphites IG-110 and IG-11 were more oxidized than medium-grained NBG-18 and NBG-17 because of deeper oxidant penetration, higher porosity and higher probability of available active sites. Variation in experimental conditions also had a marked effect on the reported kinetic parameters by several studies. Kinetic parameters such as activation energy and transition temperature were sensitive to air flow rates as well as sample size and geometry.

  12. Significance of primary irradiation creep in graphite

    CSIR Research Space (South Africa)

    Erasmus, C

    2013-05-01

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

  13. 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)

  14. 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

  15. 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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  17. Neutron irradiation effects on graphite cloth

    International Nuclear Information System (INIS)

    Gray, W.J.

    1976-01-01

    A series of cloth and fiber samples has been irradiated to fluences of 3.5, 7.3, and 10 x 10 21 cm -2 EFF* at 470 0 C. Data from the first set of samples show large shrinkages relative to that found for typical nuclear graphites. Nevertheless, all but one of the 2-dimensional cloths were unchanged except for the shrinkage. The 3-dimensional cloths, on the other hand, have deteriorated apparently because these types of weaves are less able to accommodate the large axial fiber shrinkages

  18. 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.

  19. 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.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  1. The behavior of interstitials in irradiated graphite

    International Nuclear Information System (INIS)

    Pedraza, D.F.

    1991-01-01

    A computer model is developed to simulate the behavior of self-interstitials with particular attention to clustering. Owing to the layer structure of graphite, atomistic simulations can be performed using a large parallelepipedic supercell containing a few layers. In particular, interstitial clustering is studied here using a supercell that contains two basal planes only. Frenkel pairs are randomly produced. Interstitials are placed at sites between the crystal planes while vacancies are distributed in the two crystal planes. The size of the computational cell is 20000 atoms and periodic boundary conditions are used in two dimensions. Vacancies are assumed immobile whereas interstitials are given a certain mobility. Two point defect sinks are considered, direct recombination of Frenkel pairs and interstitial clusters. The clusters are assumed to be mobile up to a certain size where they are presumed to become loop nuclei. Clusters can shrink by emission of singly bonded interstitials or by recombination of a peripheral interstitial with a neighboring vacancy. The conditions under which interstitial clustering occurs are reported. It is shown that when clustering occurs the cluster size population gradually shifts towards the largest size cluster. The implications of the present results for irradiation growth and irradiation-induced amorphization are discussed

  2. 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)

  3. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  6. 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)

  7. Reflection and photoemission studies of neutron-irradiated graphite

    International Nuclear Information System (INIS)

    Fukutani, Hirohito; Yamada, Akio; Yagi, Kazutoshi; Ooe, Satoshi; Higashiyama, Kazuyuki; Kato, Hiroo; Iwata, Tadao.

    1990-01-01

    Neutron-irradiated graphites were studied by reflectivity and photoemission (UPS, ARUPS, XPS) measurements. The π-band reflectivity peak of graphite, located at 5 eV, changed significantly and a small absorption band ascribed to vacancies produced by neutron bombardment was found to grow around 3 eV. Modification of the valence band by neutron irradiation was studied by ARUPS. The π-valence band shifts to lower binding energy towards the Fermi level and its band width becomes smaller. These results were also confirmed by the optical joint density of states obtained from K-K analysis of the reflectivity. (author)

  8. Final report on graphite irradiation test OG-3

    International Nuclear Information System (INIS)

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

    1977-01-01

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

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

    International Nuclear Information System (INIS)

    Brooking, M.

    2014-01-01

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

  10. Irradiation creep performance of graphite relevant for pebble bed HTRs

    International Nuclear Information System (INIS)

    Kleist, G.; O'Connor, M.F.

    1980-01-01

    Irradiation - induced creep in the core reflector component graphite of high temperature reactors is of primary importance to the core designer since it provides a mechanism for the relief of internal stresses arising from differential Wigner shrinkage and thermal expansion. The experimental determination of the extent of this creep for conditions relevant to the reactor is thus imperative

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  12. The irradiation creep characteristics of graphite to high fluences

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  13. Oxidation and sublimation of porous graphite during fiber laser irradiation

    Science.gov (United States)

    Phillips, Grady T.; Bauer, William A.; Gonzales, Ashley E.; Herr, Nicholas C.; Perram, Glen P.

    2017-02-01

    Porous graphite plates, cylinders and cones with densities of 1.55-1.82 g/cm3 were irradiated by a 10 kW fiber laser at 0.075 -3.525 kW/cm2 for 120 s to study mass removal and crater formation. Surface temperatures reached steady state values as high as 3767 K. The total decrease in sample mass ranged from 0.06 to 6.29 g, with crater volumes of 0.52 - 838 mm3, and penetration times for 12.7 mm thick plates as short as 38 s. Minor contaminants in the graphite samples produced calcium and iron oxide to be re-deposited on the graphite surface. Significantly increased porosity of the sample is observed even outside of the laser-irradiated region. Total mass removed increases with deposited laser energy at a rate of 4.83 g/MJ for medium extruded graphite with an apparent threshold of 0.15 MJ. Visible emission spectroscopy reveals C2 Swan and CN red, CN violet bands and Li, Na, and K 2P3/2,1/2 - 2S1/2 doublets. The reacting boundary layer is observed using a mid-wave imaging Fourier transform spectrometer (IFTS) at 2 cm-1 spectral resolution, 0.5 mm/pixel spatial resolution, and 0.75 Hz data cube rate. A two-layer radiative transfer model was used to determine plume temperature, CO, and CO2 concentrations from spectral signatures. The new understanding of graphite combustion and sublimation during laser irradiation is vital to the more complex behavior of carbon composites.

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  15. 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)

  16. Change in electrical resistance of irradiated nuclear graphite during compressive tests

    International Nuclear Information System (INIS)

    Eto, Motokuni

    1986-01-01

    In this paper the change in electrical resistance of neutron-irradiated nuclear graphite is measured and compared with that of the unirradiated; the deformation mechanism for irradiated graphite is discussed in relation to the resistance change and the defects which are believed to be produced during irradiation. (orig./RK)

  17. Graphite oxidation and damage under irradiation at high temperatures in an impure helium environment

    Science.gov (United States)

    Goodwin, Cameron S.

    The High Temperature Gas-Cooled Reactor (HTGR) is a Generation IV reactor concept that uses a graphite-moderated nuclear reactor with a once-through uranium fuel cycle. In order to investigate the mechanism for corrosion of graphite in HTGRs, the graphite was placed in a similar environment in order to evaluate its resistance to corrosion and oxidation. While the effects of radiation on graphite have been studied in the past, the properties of graphite are largely dependent on the coke used in manufacturing the graphite. There are no longer any of the previously studied graphite types available for use in the HTGR. There are various types of graphite being considered for different uses in the HTGR and all of these graphite types need to be analyzed to determine how radiation will affect them. Extensive characterization of samples of five different types of graphite was conducted. The irradiated samples were analyzed with electron paramagnetic resonance spectroscopy, Raman spectroscopy, x-ray diffraction, x-ray photoelectron spectroscopy and gas chromatography. The results prove a knowledge base for considering the graphite types best suited for use in HTGRs. In my dissertation work graphite samples were gamma irradiated and also irradiated in a mixed field, in order to study the effects of neutron as well as gamma irradiation. Thermal effects on the graphite were also investigated by irradiating the samples at room temperature and at 1000 °C. From the analysi of the samples in this study there is no evidence of substantial damage to the grades of graphite analyzed. This is significant in approving the use of these graphites in nuclear reactors. Should significant damage had occurred to the samples, the use of these grades of graphite would need to be reconsidered. This information can be used to further characterize other grades of nuclear graphite as they become available.

  18. Neutron irradiation effects on carbon and graphite cloths and fibers

    International Nuclear Information System (INIS)

    Gray, W.J.

    1977-08-01

    A series of cloth and fiber samples were irradiated to fluences of 3.5, 7.3, and 10 x 10 21 cm -2 at 470 0 C. Dimensional changes of the fibers in the radial direction ranged from -19% to +33% and in the axial direction from -18% to -27%, roughly ten times greater than dimensional changes found for typical nuclear graphites. Despite these large dimensional changes, all but one of the 2-dimensional cloths remained essentially unchanged in overall physical appearance. The 3-dimensional cloths, on the other hand, deteriorated apparently because these types of weaves were less able to accommodate the large axial fiber shrinkages

  19. Study on 14C content in post-irradiation graphite spheres of HTR-10

    International Nuclear Information System (INIS)

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

    2014-01-01

    Since the production mechanism of the 14 C in spherical fuel elements was similar to that of fuel-free graphite spheres, in order to obtain the amount of 14 C in fuel elements and graphite spheres of HTR-10, the production mechanism of the 14 C in graphite spheres was studied. The production sources of the 14 C in graphite spheres and fuel elements were summarized, the amount of 14 C in the post-irradiation graphite spheres was calculated, the decomposition techniques of graphite spheres were compared, and experimental methods for decomposing the graphite spheres and preparing the 14 C sample were proposed. The results can lay the foundation for further experimental research and provide theoretical calculations for comparison. (authors)

  20. Production of nanodiamonds by high-energy ion irradiation of graphite at room temperature

    International Nuclear Information System (INIS)

    Daulton, T.L.; Kirk, M.A.; Lewis, R.S.; Rehn, L.E.

    2001-01-01

    It has previously been shown that graphite can be transformed into diamond by MeV electron and ion irradiation at temperatures above approximately 600 deg. C. However, there exists geological evidence suggesting that carbonaceous materials can be transformed to diamond by irradiation at substantially lower temperatures. For example, submicron-size diamond aggregates have been found in uranium-rich, Precambrian carbonaceous deposits that never experienced high temperature or pressure. To test if diamonds can be formed at lower irradiation temperatures, sheets of fine-grain polycrystalline graphite were bombarded at 20 deg. C with 350±50 MeV Kr ions to fluences of 6x10 12 cm -2 using the Argonne tandem linear accelerator system (ATLAS). Ion-irradiated (and unirradiated control) graphite specimens were then subjected to acid dissolution treatments to remove untransformed graphite and isolate diamonds that were produced; these acid residues were subsequently characterized by high-resolution and analytical electron microscopy. The acid residue of the ion-irradiated graphite was found to contain nanodiamonds, demonstrating that ion irradiation of graphite at ambient temperature can produce diamond. The diamond yield under our irradiation conditions is low, ∼0.01 diamonds/ion. An important observation that emerges from comparing the present result with previous observations of diamond formation during irradiation is that nanodiamonds form under a surprisingly wide range of irradiation conditions. This propensity may be related to the very small difference in the graphite and diamond free-energies coupled with surface-energy considerations that may alter the relative stability of diamond and graphite at nanometer sizes

  1. 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)

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

    International Nuclear Information System (INIS)

    Rouaud, Michel

    1969-01-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    Bocquet, M.; Micaud, G.

    1962-01-01

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

  5. LIMITED OXIDATION OF IRRADIATED GRAPHITE WASTE TO REMOVE SURFACE CARBON-14

    Directory of Open Access Journals (Sweden)

    TARA E. SMITH

    2013-04-01

    Full Text Available Large quantities of irradiated graphite waste from graphite-moderated nuclear reactors exist and are expected to increase in the case of High Temperature Reactor (HTR deployment [1,2]. This situation indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 (14C, with a half-life of 5730 years. Fachinger et al. [2] have demonstrated that thermal treatment of irradiated graphite removes a significant fraction of the 14C, which tends to be concentrated on the graphite surface. During thermal treatment, graphite surface carbon atoms interact with naturally adsorbed oxygen complexes to create COx gases, i.e. “gasify” graphite. The effectiveness of this process is highly dependent on the availability of adsorbed oxygen compounds. The quantity and form of adsorbed oxygen complexes in pre- and post-irradiated graphite were studied using Time of Flight Secondary Ion Mass Spectrometry (ToF-SIMS and X-ray Photoelectron Spectroscopy (XPS in an effort to better understand the gasification process and to apply that understanding to process optimization. Adsorbed oxygen fragments were detected on both irradiated and unirradiated graphite; however, carbon-oxygen bonds were identified only on the irradiated material. This difference is likely due to a large number of carbon active sites associated with the higher lattice disorder resulting from irradiation. Results of XPS analysis also indicated the potential bonding structures of the oxygen fragments removed during surface impingement. Ester- and carboxyl- like structures were predominant among the identified oxygen-containing fragments. The indicated structures are consistent with those characterized by Fanning and Vannice [3] and later incorporated into an oxidation kinetics model by El-Genk and Tournier [4]. Based on the predicted desorption mechanisms of carbon oxides from the identified compounds, it is expected that a

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  7. Processing of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal. Results of a Coordinated Research Project

    International Nuclear Information System (INIS)

    2016-05-01

    Graphite is widely used in the nuclear industry and in research facilities and this has led to increasing amounts of irradiated graphite residing in temporary storage facilities pending disposal. This publication arises from a coordinated research project (CRP) on the processing of irradiated graphite to meet acceptance criteria for waste disposal. It presents the findings of the CRP, the general conclusions and recommendations. The topics covered include, graphite management issues, characterization of irradiated graphite, processing and treatment, immobilization and disposal. Included on the attached CD-ROM are formal reports from the participants

  8. 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

  9. Deformation and fracture of irradiated polygranular pile grade A reactor core graphite

    International Nuclear Information System (INIS)

    Heard, P.J.; Wootton, M.R.; Moskovic, R.; Flewitt, P.E.J.

    2011-01-01

    Highlights: → Mechanical properties of PGA nuclear graphite specimens were tested. → Load-displacement characteristics were consistent with quasi-brittle behaviour. → Micro-cracks observed in non-linear part of load-displacement curve pre-peak load. → Micro-testing showed surface tilts consistent with twinning. → Irradiated specimens failed at 20% lower load for 15% weight loss material. - Abstract: Pile grade A (PGA) graphite is used as a moderator in UK gas cooled nuclear reactors. This is a polygranular, aggregate material with quasi-brittle behaviour. When exposed to the service environment the material is subject to radiolytic oxidation that results in mass loss and an attendant increase in porosity. In the present work both unirradiated and irradiated small specimens of PGA graphite have been subjected to diametral compression. A novel trench-probe loading method is also described that allows micro-scale specimens prepared by focused ion beam milling to be fractured in a focused ion beam work station. This allows the fracture characteristics of selected regions of the graphite microstructure to be interrogated. The load-displacement and fracture characteristics of both the unirradiated and irradiated PGA graphite are compared and shown to be consistent with quasi-brittle behaviour. In addition, surface features consistent with elastically induced twins are observed associated with filler particles of the graphite. The results are discussed with respect to the quasi-brittle behaviour of this polygranular graphite.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

    Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR) which is a graphite-moderated and helium gas-cooled reactor. In the case of air ingress accident in HTGR, SiO 2 protective layer is formed on the surface of SiC layer in TRISO CFP and oxidation of SiC does not proceed and fission products are retained inside the fuel particle. A new safety concept for the HTGR, called Naturally Safe HTGR, has been recently proposed. To enhance the safety of Naturally Safe HTGR ultimately, it is expected that oxidation-resistant graphite is used for graphite components to prevent the TRISO CFPs and fuel compacts from failure. SiC coating is one of candidate methods for oxidation-resistant graphite. JAEA and four graphite companies launched R&Ds to develop the oxidation-resistant graphite and the International Science and Technology Center (ISTC) partner project with JAEA and INP was launched to investigate the irradiation effects on the oxidation-resistant graphite. To determine grades of the oxidation-resistant graphite which will be adopted as irradiation test, a preliminary oxidation test was carried out. This paper described the results of the preliminary oxidation test, the plan of out-of-pile test, irradiation test and post-irradiation test (PIE) of the oxidation-resistant graphite. The results of the preliminary oxidation test showed that the integrity of the oxidation resistant graphite was confirmed and that all of grades used in the preliminary test can be adopted as the irradiation test. Target irradiation temperature was determined to be 1473 (K) and neutron fluence was determined to be from 0.54 × 10 25 through 1.4 × 10 25 (/m 2 , E>0.18MeV). Weight change, oxidation rate, activation energy, surface condition, etc. will be evaluated in out-of-pile test and weight change, irradiation effect on oxidation rate and activation energy, surface condition, etc. will be evaluated in PIE. (author)

  11. A review of irradiation induced creep in graphite under CAGR conditions

    International Nuclear Information System (INIS)

    Brocklehurst, J.E.; Kelly, B.T.

    1989-06-01

    Graphite irradiation induced creep data have been reviewed in detail and compared with the existing model used for stress calculations under Commercial Advanced Gas Cooled Reactor conditions. The relationship between creep and elastic modulus is well established and allows the creep behaviour of any graphite in any orientation to be predicted. The model predicts the initial build up of creep strain in different graphites extremely well. However, there are differences between prediction and experiment in creep at high doses; a creep test on a pre-irradiated specimen showed rather more creep ductility than predicted, whilst in experiments under a constant stress applied for the whole period of irradiation, the creep rates decreased to lower values than predicted. It is suggested that to obtain better estimates of brick stresses future irradiation experiments should aim to generate creep data under realistic variations in applied stress level, or rather creep strain history. (author)

  12. Irradiation of graphite cloth at various temperatures with deutrons and helium ions

    International Nuclear Information System (INIS)

    Ekern, R.; Das, S.K.; Kaminsky, M.

    1975-01-01

    Graphite cloth samples were irradiated with 100 keV deuterons and 4 He + ions at room temperature and at elevated temperatures. Scanning electron microscopy was used to examine the surfaces of irradiated and unirradiated graphite fibers. Irradiation at room temperature with 4 He + to a total dose of 3.1 x 10 18 ions cm -2 produces considerable flaking of individual fibers, which is not observed on unirradiated fibers. Identical irradiations at 400 0 and 800 0 with 4 He + did not produce any detectable flaking or other surface damage. The elevated temperatures apparently prevent an accumulation of helium in localized areas which in turn could cause flaking in near surface regions. Results obtained for deuteron bombardment of graphite cloth at room temperature and at 600 0 C are also discussed

  13. Tests for removal of Co-60 and Eu-154 from irradiated graphite in the TRIGA Reactor

    International Nuclear Information System (INIS)

    Arsene, Carmen

    2009-01-01

    The irradiated graphite in Romania is mainly generated in the thermal columns of TRIGA and WWER-S research reactors (about 9 tones). It was found that the radionuclide content of the graphite irradiated in the TRIGA research reactor is mainly due to C-14 (103 Bq/g), Eu-152 (600-700 Bq/g) and Co-60 (130-150 Bq/g) and low amounts of Eu-154 and Cs-137, depending on location in the thermal column and on irradiation history. In order to minimize the waste inventory and volume in view of their final disposal, in the present paper we show the results of experiments performed for developing and optimizing methods for the chemical decontamination of the irradiated graphite. These procedures are based on strong alkaline solutions for Eu-152 and strong acid solutions for Co-60. The influence of the process parameters on the decontamination factor is investigated. (authors)

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

    International Nuclear Information System (INIS)

    Matsuo, Hideto

    1988-01-01

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

  15. 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)

  16. 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.

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

    Directory of Open Access Journals (Sweden)

    Xuegang Liu

    2017-01-01

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

  18. Updating irradiated graphite disposal: Project 'GRAPA' and the international decommissioning network.

    Science.gov (United States)

    Wickham, Anthony; Steinmetz, Hans-Jürgen; O'Sullivan, Patrick; Ojovan, Michael I

    2017-05-01

    Demonstrating competence in planning and executing the disposal of radioactive wastes is a key factor in the public perception of the nuclear power industry and must be demonstrated when making the case for new nuclear build. This work addresses the particular waste stream of irradiated graphite, mostly derived from reactor moderators and amounting to more than 250,000 tonnes world-wide. Use may be made of its unique chemical and physical properties to consider possible processing and disposal options outside the normal simple classifications and repository options for mixed low or intermediate-level wastes. The IAEA has an obvious involvement in radioactive waste disposal and has established a new project 'GRAPA' - Irradiated Graphite Processing Approaches - to encourage an international debate and collaborative work aimed at optimising and facilitating the treatment of irradiated graphite. Copyright © 2017 Elsevier Ltd. All rights reserved.

  19. 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

  20. An analytical study on porosity changes of nuclear graphites under high temperature irradiations

    International Nuclear Information System (INIS)

    Arai, T.

    1996-01-01

    A quantitative description of the changing pore structure, based on some radiation damage mechanisms, may introduce a physically appropriate method for lifetime assessment of graphite fuel and moderator components. Recently Brocklehurst and Kelly have analyzed well-characterized data on dimensional changes of UK reactor graphites to quantify volumetric and linear pore generation terms. The analysis (B/K theory) has demonstrated that a crystal strain parameter X T , depending on irradiation temperature and fluence, is suitable for defining structure factors, which relate changes in microstructure with those in macroscopic properties of a family of nuclear graphites. Graphite components in high temperature reactors are subjected to higher temperatures well above 1000 deg. C, which accelerate pore generation. Their mechanical integrity will suffer from the deterioration, resulting in a reduced lifetime. Previous design considerations on the dimensional change behavior have been based on an empirical approach using measured data obtained in a number of irradiation experiments. A large variety of experimental data have been utilized to develop a general phenomenological model(Graphite Damage Model, GDM) for predicting engineering properties of nuclear graphites. The present study tries to combine the B/K theory with the GDM prediction with a view to characterizing porosity changes at high temperatures of some graphites from different manufacturing routes. The dimensional change data in the literature are analyzed by the GDM to obtain their analytical presentation as a function of temperature and fluence. The results are used to derive an X T function and pore volume change as a function of X T for each grade of graphite. The resulting porosity changes are compared between different kinds of graphites. 13 refs, 6 figs, 3 tabs

  1. Mass removal by oxidation and sublimation of porous graphite during fiber laser irradiation

    Science.gov (United States)

    Phillips, Grady T.; Bauer, William A.; Fox, Charles D.; Gonzales, Ashley E.; Herr, Nicholas C.; Gosse, Ryan C.; Perram, Glen P.

    2017-01-01

    The various effects of laser heating of carbon materials are key to assessing laser weapon effectiveness. Porous graphite plates, cylinders, and cones with densities of 1.55 to 1.82 g/cm3 were irradiated by a 10-kW fiber laser at 0.075 to 3.525 kW/cm2 for 120 s to study mass removal and crater formation. Surface temperatures reached steady state values as high as 3767 K. The total decrease in sample mass ranged from 0.06 to 6.29 g, with crater volumes of 0.52 to 838 mm3, and penetration times for 12.7-mm-thick plates as short as 38 s. Minor contaminants in the graphite samples produced calcium and iron oxide to be redeposited on the graphite surface. Dramatic graphite crystalline structures are also produced at higher laser irradiances. Significantly increased porosity of the sample is observed even outside the laser-irradiated region. Total mass removed increases with deposited laser energy at a rate of 4.83 g/MJ for medium extruded graphite with an apparent threshold of 0.15 MJ. At ˜3.5 kW/cm2, the fractions of the mass removed from the cylindrical samples in the crater, surrounding trench, and outer region of decreased porosity are 38%, 47%, and 15%, respectively. Graphite is particularly resistant to damage by high power lasers. The new understanding of graphite combustion and sublimation during laser irradiation is vital to the more complex behavior of carbon composites.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-12-01

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

  3. Swift heavy ions induced irradiation effects in monolayer graphene and highly oriented pyrolytic graphite

    International Nuclear Information System (INIS)

    Zeng, J.; Yao, H.J.; Zhang, S.X.; Zhai, P.F.; Duan, J.L.; Sun, Y.M.; Li, G.P.; Liu, J.

    2014-01-01

    Monolayer graphene and highly oriented pyrolytic graphite (HOPG) were irradiated by swift heavy ions ( 209 Bi and 112 Sn) with the fluence between 10 11 and 10 14 ions/cm 2 . Both pristine and irradiated samples were investigated by Raman spectroscopy. It was found that D and D′ peaks appear after irradiation, which indicated the ion irradiation introduced damage both in the graphene and graphite lattice. Due to the special single atomic layer structure of graphene, the irradiation fluence threshold Φ th of the D band of graphene is significantly lower ( 11 ions/cm 2 ) than that (2.5 × 10 12 ions/cm 2 ) of HOPG. The larger defect density in graphene than in HOPG indicates that the monolayer graphene is much easier to be damaged than bulk graphite by swift heavy ions. Moreover, different defect types in graphene and HOPG were detected by the different values of I D /I D′ . For the irradiation with the same electronic energy loss, the velocity effect was found in HOPG. However, in this experiment, the velocity effect was not observed in graphene samples irradiated by swift heavy ions

  4. 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

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

    Directory of Open Access Journals (Sweden)

    N. Simos

    2017-07-01

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

  6. Annealing of neutron damage in graphite irradiated and stored at room temperature

    International Nuclear Information System (INIS)

    Gray, W.J.; Thrower, P.A.

    1979-01-01

    Annealing of neutron radiation damage in graphite at the same temperature at which it was irradiated is reported. Highly oriented pyrolytic graphite samples were irradiated to fluences in the range 0.44 to 153 x 10 15 /cm 2 at room temperature using three different neutron sources with average energies of 1.5, 5.5, and 15 MeV, respectively. Following these irradiations, the C 44 elastic constants of the samples were measured several times over periods up to two years during which time sample temperatures never exceeded 30 0 C. The C 44 constants were observed to slowly decrease toward their unirradiated values with up to 40% of the irradiation-induced changes eventually annealing out

  7. Rapid analysis method for the determination of 14C specific activity in irradiated graphite.

    Science.gov (United States)

    Remeikis, Vidmantas; Lagzdina, Elena; Garbaras, Andrius; Gudelis, Arūnas; Garankin, Jevgenij; Plukienė, Rita; Juodis, Laurynas; Duškesas, Grigorijus; Lingis, Danielius; Abdulajev, Vladimir; Plukis, Artūras

    2018-01-01

    14C is one of the limiting radionuclides used in the categorization of radioactive graphite waste; this categorization is crucial in selecting the appropriate graphite treatment/disposal method. We propose a rapid analysis method for 14C specific activity determination in small graphite samples in the 1-100 μg range. The method applies an oxidation procedure to the sample, which extracts 14C from the different carbonaceous matrices in a controlled manner. Because this method enables fast online measurement and 14C specific activity evaluation, it can be especially useful for characterizing 14C in irradiated graphite when dismantling graphite moderator and reflector parts, or when sorting radioactive graphite waste from decommissioned nuclear power plants. The proposed rapid method is based on graphite combustion and the subsequent measurement of both CO2 and 14C, using a commercial elemental analyser and the semiconductor detector, respectively. The method was verified using the liquid scintillation counting (LSC) technique. The uncertainty of this rapid method is within the acceptable range for radioactive waste characterization purposes. The 14C specific activity determination procedure proposed in this study takes approximately ten minutes, comparing favorably to the more complicated and time consuming LSC method. This method can be potentially used to radiologically characterize radioactive waste or used in biomedical applications when dealing with the specific activity determination of 14C in the sample.

  8. Rapid analysis method for the determination of 14C specific activity in irradiated graphite.

    Directory of Open Access Journals (Sweden)

    Vidmantas Remeikis

    Full Text Available 14C is one of the limiting radionuclides used in the categorization of radioactive graphite waste; this categorization is crucial in selecting the appropriate graphite treatment/disposal method. We propose a rapid analysis method for 14C specific activity determination in small graphite samples in the 1-100 μg range. The method applies an oxidation procedure to the sample, which extracts 14C from the different carbonaceous matrices in a controlled manner. Because this method enables fast online measurement and 14C specific activity evaluation, it can be especially useful for characterizing 14C in irradiated graphite when dismantling graphite moderator and reflector parts, or when sorting radioactive graphite waste from decommissioned nuclear power plants. The proposed rapid method is based on graphite combustion and the subsequent measurement of both CO2 and 14C, using a commercial elemental analyser and the semiconductor detector, respectively. The method was verified using the liquid scintillation counting (LSC technique. The uncertainty of this rapid method is within the acceptable range for radioactive waste characterization purposes. The 14C specific activity determination procedure proposed in this study takes approximately ten minutes, comparing favorably to the more complicated and time consuming LSC method. This method can be potentially used to radiologically characterize radioactive waste or used in biomedical applications when dealing with the specific activity determination of 14C in the sample.

  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. 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)

  11. Spectroscopic study of energetic helium-ion irradiation effects on nuclear graphite tiles

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Wan; Lee, K.W. [Department of Physics, Korea University, Seoul 136-713 (Korea, Republic of); Choi, D.M.; Noh, S.J.; Kim, H.S. [Department of Applied Physics, Dankook University, Yongin 448-701 (Korea, Republic of); Lee, Cheol Eui, E-mail: rscel@korea.ac.kr [Department of Physics, Korea University, Seoul 136-713 (Korea, Republic of)

    2016-02-01

    Highlights: • Energetic helium-ion irradiation on nuclear graphite tiles studied for plasma facing components. • XPS reveals recrystallization at low dose irradiation and DLC sites at higher doses. • Raman spectroscopy reveals increasing diamond-like defects and structural deformation. • Average inter-defect distance obtained as a function of irradiation dose from Raman intensities. - Abstract: Helium ion-irradiation effects on the nuclear graphite tiles were studied in order to understand the structural modifications and damages that can be produced by fusion reaction in tokamaks. The surface morphological changes due to increasing dose of the irradiation were examined by the field-effect scanning electron microscopy, and X-ray photoelectron spectroscopy elucidated the changes in the shallow surface bonding configurations caused by the energetic irradiation. Raman spectroscopy revealed the structural defects and diamond-like carbon sites that increased with increasing irradiation dose, and the average inter-defect distance was found from the Raman peak intensities as a function of the irradiation dose.

  12. Quenchable compressed graphite synthesized from neutron-irradiated highly oriented pyrolytic graphite in high pressure treatment at 1500 °C

    Science.gov (United States)

    Niwase, Keisuke; Terasawa, Mititaka; Honda, Shin-ichi; Niibe, Masahito; Hisakuni, Tomohiko; Iwata, Tadao; Higo, Yuji; Hirai, Takeshi; Shinmei, Toru; Ohfuji, Hiroaki; Irifune, Tetsuo

    2018-04-01

    The super hard material of "compressed graphite" (CG) has been reported to be formed under compression of graphite at room temperature. However, it returns to graphite under decompression. Neutron-irradiated graphite, on the other hand, is a unique material for the synthesis of a new carbon phase, as reported by the formation of an amorphous diamond by shock compression. Here, we investigate the change of structure of highly oriented pyrolytic graphite (HOPG) irradiated with neutrons to a fluence of 1.4 × 1024 n/m2 under static pressure. The neutron-irradiated HOPG sample was compressed to 15 GPa at room temperature and then the temperature was increased up to 1500 °C. X-ray diffraction, high-resolution transmission electron microscopy on the recovered sample clearly showed the formation of a significant amount of quenchable-CG with ordinary graphite. Formation of hexagonal and cubic diamonds was also confirmed. The effect of irradiation-induced defects on the synthesis of quenchable-CG under high pressure and high temperature treatment was discussed.

  13. 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

  14. Influence of neutron irradiation on the microstructure of nuclear graphite: An X-ray diffraction study

    Science.gov (United States)

    Zhou, Z.; Bouwman, W. G.; Schut, H.; van Staveren, T. O.; Heijna, M. C. R.; Pappas, C.

    2017-04-01

    Neutron irradiation effects on the microstructure of nuclear graphite have been investigated by X-ray diffraction on virgin and low doses (∼ 1.3 and ∼ 2.2 dpa), high temperature (750° C) irradiated samples. The diffraction patterns were interpreted using a model, which takes into account the turbostratic disorder. Besides the lattice constants, the model introduces two distinct coherent lengths in the c-axis and the basal plane, that characterise the volumes from which X-rays are scattered coherently. The methodology used in this work allows to quantify the effect of irradiation damage on the microstructure of nuclear graphite seen by X-ray diffraction. The results show that the changes of the deduced structural parameters are in agreement with previous observations from electron microscopy, but not directly related to macroscopic changes.

  15. A Review of Graphite Irradiation Creep Data from the "OC-Series" of Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Davies, Mark A. [MARAD Co. Ltd., Washington, DC (United States); Burchell, Timothy D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2012-09-01

    The OC-Series graphite irradiation creep experiments were conducted in the early 1970s in the Oak Ridge Research Reactor (ORR) at ORNL. The OC Series consisted of 5 experiments, Capsules 1, 3 and 5 were irradiated at 900°C and Capsules 2 and 4 were irradiated at 600°C. Each capsule contained four columns of specimens, two loaded in compression and two un-loaded. The loaded columns had specimens of different diameter to generate two stress levels, 13.8 MPa and 20.7 MPa. Some of the data from these experiments were presented in extended abstracts at a Carbon Conference (Kennedy et al, 1977: Kennedy and Eatherly, 1979). The data presented some challenges to the accepted approach to irradiation induced creep in graphite adopted in the UK, specifically lateral creep strain behaviour and the effect of irradiation induced creep strain on material properties, e.g. CTE and Poisson’s Ratio. A recent review of irradiation induced creep (Davies & Bradford, 2004) included an anlaysis of the available OC-series data (Mobasheran, 1990) and led to a request to ORNL for an examination of the original OC-Series dataset. An initial search of the ORNL archive revealed additional data from the OC-Series experiment including previously unknown irradiation annealing experiments. This report presents a re-analysis of the available data from the OC-Series archive.

  16. Erosion of CFC, pyrolytic and boronated graphite under short pulsed laser irradiation

    International Nuclear Information System (INIS)

    Kraaij, G.J.; Bakker, J.; Stad, R.C.L. van der

    1992-07-01

    The effect of short pulsed laser irradiation of '0/3' ms and up to 10 MJ/m 2 on different types of carbon base materials is described. These materials are investigated as candidate protection materials for the Plasma Facing Components of NET/ITER. These materials are: carbon fibre composite graphite, pyrolytic graphite and boronated graphite. The volume of the laser induced craters was measured with an optical topographic scanner, and these data are evaluated with a simple model for the erosion. As a results, the enthalpy of ablation is estimated as 30±3 MJ/kg. A comparison is made with finite element numerical calculations, and the effect of lateral heat transfer is estimated using an analytical model. (author). 8 refs., 23 figs., 4 tabs

  17. Study on structural recovery of graphite irradiated with swift heavy ions at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Pellemoine, F., E-mail: pellemoi@frib.msu.edu [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Avilov, M. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Bender, M. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Ewing, R.C. [Dept. of Geological Sciences, Stanford University, Stanford, CA 94305-2115 (United States); Fernandes, S. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Lang, M. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996-2300 (United States); Li, W.X. [Dept. of Geological Sciences, Stanford University, Stanford, CA 94305-2115 (United States); Mittig, W. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); National Superconducting Cyclotron Laboratory, Michigan State University, East Lansing, MI 48824-1321 (United States); Schein, M. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Severin, D. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Tomut, M. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Laboratory of Magnetism and Superconductivity, National Institute for Materials Physics NIMP, Bucharest (Romania); Trautmann, C. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Dept. of Materials Science, Technische Universität Darmstadt, Darmstadt (Germany); and others

    2015-12-15

    Thin graphite foils bombarded with an intense high-energy (8.6 MeV/u) gold beam reaching fluences up to 1 × 10{sup 15} ions/cm{sup 2} lead to swelling and electrical resistivity changes. As shown earlier, these effects are diminished with increasing irradiation temperature. The work reported here extends the investigation of beam induced changes of these samples by structural analysis using synchrotron X-ray diffraction and transmission electron microscope. A nearly complete recovery from swelling at irradiation temperatures above about 1500 °C is identified.

  18. Micro to nanostructural observations in neutron irradiated nuclear graphites PCEA and PCIB

    Science.gov (United States)

    Freeman, H. M.; Mironov, B. E.; Windes, W.; Alnairi, M. M.; Scott, A. J.; Westwood, A. V. K.; Brydson, R. M. D.

    2017-08-01

    The neutron irradiation-induced structural changes in nuclear grade graphites PCEA and PCIB were investigated using scanning electron microscopy (SEM), X-ray diffraction (XRD), Raman spectroscopy, transmission electron microscopy (TEM), selected area electron diffraction (SAED) and electron energy loss spectroscopy (EELS). The graphite samples were irradiated at the Advanced Test Reactor at the Idaho National Laboratory. Received doses ranged from 1.5 to 6.8 displacements per atom and irradiation temperatures varied between 350 °C and 670 °C. XRD and Raman measurements provided evidence for irradiation induced crystallite fragmentation, with crystallite sizes reduced by 39-55%. Analysis of TEM images was used to quantify fringe length, tortuosity, and relative misorientation of planes, and indicated that neutron irradiation induced basal plane fragmentation and curvature. EELS was used to quantify the proportion of sp2 bonding and specimen density; a slight reduction in planar-sp2 content (due to the buckling basal planes and the introduction of non-six-membered rings) agreed with the observations from TEM.

  19. 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

  20. 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.

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

    Science.gov (United States)

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

    2015-12-01

    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.

  2. Processing of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal. Results of a Coordinated Research Project. Companion CD-ROM

    International Nuclear Information System (INIS)

    2016-05-01

    Graphite is widely used in the nuclear industry and in research facilities and this has led to increasing amounts of irradiated graphite residing in temporary storage facilities pending disposal. This publication arises from a coordinated research project (CRP) on the processing of irradiated graphite to meet acceptance criteria for waste disposal. It presents the findings of the CRP, the general conclusions and recommendations. The topics covered include, graphite management issues, characterization of irradiated graphite, processing and treatment, immobilization and disposal. Included on the attached CD-ROM are formal reports from the participants

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  4. Simultaneous heating and compression of irradiated graphite during synchrotron microtomographic imaging

    Science.gov (United States)

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

    2017-06-01

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

  5. 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...

  6. Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Grover, S. Blaine

    2009-01-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or

  7. Irradiation effects on thermal shock resistance and its fracture toughness of HTGR graphites

    International Nuclear Information System (INIS)

    Sato, Sennosuke; Imamura, Yoshio; Kawamata, Kiyohiro; Awaji, Hideo; Oku, Tatsuo.

    1979-01-01

    This paper describes changes in the thermal shock resistance delta = σsub(t)k / E alpha (σsub(t): tensile strength, k: thermal conductivity, E: Young's modulus, alpha : coefficient of thermal expansion) and the thermal shock fracture toughness delta = K sub( ic)k / E alpha (K sub( ic): fracture toughness value of the mode I) in addition to usual mechanical properties including the diametral compressive strength and fracture toughness of four varieties of graphite (IM2-24, 7477, H327 and SMG) for the high temperature gas-cooled reactor due to neutron irradiations of (1.6 -- 2.3) x 10 21 n/cm 2 ( gt 0.18 MeV) at 600 -- 850 0 C. These experiments are carried out by means of our recently developed techniques using small disk type specimens which are very effective for a capsule irradiation in the JMTR. Both the thermal shock resistance and the thermal shock fracture toughness of graphites after irradiation are expressed to decrease remarkably in contrast with the increase of the usual mechanical strength. (author)

  8. Transient analysis of nuclear graphite oxidation for high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Wei, E-mail: wxu12@mails.tsinghua.edu.cn; Shi, Lei; Zheng, Yanhua

    2016-09-15

    Graphite is widely used as moderator, reflector and structural materials in the high temperature gas-cooled reactor pebble-bed modular (HTR-PM). In normal operating conditions or water/air ingress accident, the nuclear graphite in the reactor may be oxidized by air or steam. Oxidation behavior of nuclear graphite IG-110 which is used as the structural materials and reflector of HTR-PM is mainly researched in this paper. To investigate the penetration depth of oxygen in IG-110, this paper developed the one dimensional spherical oxidation model. In the oxidation model, the equations considered graphite porosity variation with the graphite weight loss. The effect of weight loss on the effective diffusion coefficient and the oxidation rate was also considered in this model. Based on this theoretical model, this paper obtained the relative concentration and local weight loss ratio profile in graphite. In addition, the local effective diffusion coefficient and oxidation rate in the graphite were also investigated.

  9. 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)

  10. 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)

  11. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-10

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

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

    International Nuclear Information System (INIS)

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

    1984-05-01

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

  14. Tritium retention in neutron irradiated graphites, CFC's and doped C composites

    International Nuclear Information System (INIS)

    Kwast, H.; Muis, R.P.; Boshoven, J.G.

    1993-04-01

    Experiments have been performed to determine the conditions for loading graphite samples with tritium by exposure to a tritium containing gas. This is necessary to investigate the retention of tritium in irradiated graphite. The influence of exposure temperature, pressure and time as well as the sample size has been investigated. The exposure temperatures ranged from 750 C to 1150 C. The exposure time was varied from 0 to 24 hours, while total pressures of about 700 mbar to about 1000 mbar were used. So far the exposure gas was helium with 100 ppm H 2 and contained about 2.5 μCi/ml tritium at NTP, which is about 1 vppm. The effect of composition of the exposure gas and tritium concentration has still to be determined. Also longer exposure times than 24 h have to be applied to see whether or not saturation will occur at 850 C. The tritium retention of the unirradiated materials S 1611, A 05, CL 5890, Pfizer Pyrolitic and Sepcarb N112 were 18.5, 11.6, 7.8, 7.8 and 3.2 μCi/g, respectively. The exposure conditions were: 850 C, 10 hours and 800 mbar, using He/H 2 /T 2 as exposure gas. Ot the tested materials Sepcarb N112 has the lowest retention of tritium. (orig.)

  15. Surface structure modification of single crystal graphite after slow, highly charged ion irradiation

    Science.gov (United States)

    Alzaher, I.; Akcöltekin, S.; Ban-d'Etat, B.; Manil, B.; Dey, K. R.; Been, T.; Boduch, P.; Rothard, H.; Schleberger, M.; Lebius, H.

    2018-04-01

    Single crystal graphite was irradiated by slow, highly charged ions. The modification of the surface structure was studied by means of Low-Energy Electron Diffraction. The observed damage cross section increases with the potential energy, i.e. the charge state of the incident ion, at a constant kinetic energy. The potential energy is more efficient for the damage production than the kinetic energy by more than a factor of twenty. Comparison with earlier results hints to a strong link between early electron creation and later target atom rearrangement. With increasing ion fluence, the initially large-scale single crystal is first transformed into μ m-sized crystals, before complete amorphisation takes place.

  16. Investigation of corrosion resistance of graphite under electron irradiation in the oxygen flow at the temperatures 600...800 deg C

    International Nuclear Information System (INIS)

    Zelenskij, V.F.; Odejchuk, N.P.; Ryzhov, V.P.; Borisenko, V.N.; Gamov, V.O.; Lyashchenko, A.N.; Ulybkin, A.L.; Yakovlev, V.K.

    2013-01-01

    In work results of researches of corrosion resistance of graphite samples by grades MPG, ARV and GSP (graphite bonded pyrocarbon) in oxygen flow at the temperatures of ∼ 600 and ∼ 800 deg C under the influence of electron irradiation at the accelerator ELIAS. Established that the oxidation process of graphite with the increasing temperature goes significantly more intensively and the oxidation rate increases in 6...8 times. It is shown that the best corrosion resistance under irradiation in the investigated temperature range has graphite GSP with density 1.77...1.9 g/cm 3 manufacturing of NSC KIPT

  17. 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)

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

    Science.gov (United States)

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

    2018-04-01

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

  19. The effects of neutron irradiation on various fuels used in gas graphite reactor systems

    International Nuclear Information System (INIS)

    Englander, M.

    1964-01-01

    The behaviour of natural uranium based fuels in form of rods and tubes has been examined after irradiation in several French reactors. Two categories of uranium alloys have thus been studied: on one hand alloys containing from 0,5 to 3 p.100 by weight of molybdenum, and on the other hand alloys obtained by addition to nuclear purity uranium, of less than 1500 ppm of metallic elements such as iron, aluminium, chromium and silicon. The tests have been carried out with two types of elements: one canned with a magnesium alloy and cooled by carbon dioxide under pressure (in G 2 G 3), the other canned with aluminium and cooled by heavy water (in EL 3). Each of these elements had the same outside form as the usual elements in these reactors, but the enrichment and thickness of the fuel were adapted in order to meet different irradiation conditions. The post-irradiation examinations described, thereafter involved techniques of metrology, of radiography, of density measurements before and after isochron annealings, as well as macro-graphies or optical and electronic micrographies. The experimental results thus obtained reflect the changes in the internal and external structures of the full-size fuel elements; they are discussed in terms of the combined effects of pressure, temperature and burn-up on the morphological state, on the isotropy and on the mechanical characteristics of the initial polycrystalline aggregate. An analysis of the results makes it possible to deduce the various advantages of un-enriched metallic fuels designed to operate with a hot-point of around 600 C in reactors of the gas-graphite type, and for a length of time corresponding to an output of 4500 MW day/metre ton. (author) [fr

  20. 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

  1. Deuterium migration in nuclear graphite: Consequences for the behavior of tritium in CO2-cooled reactors and for the decontamination of irradiated graphite waste

    Science.gov (United States)

    Le Guillou, M.; Toulhoat, N.; Pipon, Y.; Moncoffre, N.; Khodja, H.

    2015-06-01

    In this paper, we aim at understanding tritium behavior in the graphite moderator of French CO2-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+ 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 CO2) 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 total amount produced

  2. Microstructural Characterization of Next Generation Nuclear Graphites

    Energy Technology Data Exchange (ETDEWEB)

    Karthik Chinnathambi; Joshua Kane; Darryl P. Butt; William E. Windes; Rick Ubic

    2012-04-01

    This article reports the microstructural characteristics of various petroleum and pitch based nuclear graphites (IG-110, NBG-18, and PCEA) that are of interest to the next generation nuclear plant program. Bright-field transmission electron microscopy imaging was used to identify and understand the different features constituting the microstructure of nuclear graphite such as the filler particles, microcracks, binder phase, rosette-shaped quinoline insoluble (QI) particles, chaotic structures, and turbostratic graphite phase. The dimensions of microcracks were found to vary from a few nanometers to tens of microns. Furthermore, the microcracks were found to be filled with amorphous carbon of unknown origin. The pitch coke based graphite (NBG-18) was found to contain higher concentration of binder phase constituting QI particles as well as chaotic structures. The turbostratic graphite, present in all of the grades, was identified through their elliptical diffraction patterns. The difference in the microstructure has been analyzed in view of their processing conditions.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  4. Irradiation testing of internally pressurized and/or graphite coated Zircaloy-4 clad fuel rods in the NRX Reactor (AWBA Development Program)

    International Nuclear Information System (INIS)

    Hoffman, R.C.; Sherman, J.

    1978-11-01

    Irradiation tests on 0.612 inch O.D. by 117-inch long Zircaloy-4 clad fuel rods were performed to assess the effects on fuel rod performance of (1) internal helium pre-pressurization to 500 psi as fabricated, (2) the presence of a graphite barrier coating on the inside cladding surface, and (3) combined pre-pressurization and graphite coating. Periodic dimensional examinations were performed on the test rods, and the results were compared with data obtained from two previously irradiated test rods--both unpressurized and uncoated and one intentionally defected. These comparisons indicate that both pre-pressurization and graphite coating can substantially improve fuel element performance capability

  5. Radiation damage in graphite

    CERN Document Server

    Simmons, John Harry Walrond

    1965-01-01

    Nuclear Energy, Volume 102: Radiation Damage in Graphite provides a general account of the effects of irradiation on graphite. This book presents valuable work on the structure of the defects produced in graphite crystals by irradiation. Organized into eight chapters, this volume begins with an overview of the description of the methods of manufacturing graphite and of its physical properties. This text then presents details of the method of setting up a scale of irradiation dose. Other chapters consider the effect of irradiation at a given temperature on a physical property of graphite. This

  6. Development of a Scanning Microscale Fast Neutron Irradiation Platform for Examining the Correlation Between Local Neutron Damage and Graphite Microstructure

    Energy Technology Data Exchange (ETDEWEB)

    Pinhero, Patrick [Univ. of Missouri, Columbia, MO (United States); Windes, William [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-10

    The fast particle radiation damage effect of graphite, a main material in current and future nuclear reactors, has significant influence on the utilization of this material in fission and fusion plants. Atoms on graphite crystals can be easily replaced or dislocated by fast protons and result in interstitials and vacancies. The currently accepted model indicates that after most of the interstitials recombine with vacancies, surviving interstitials form clusters and furthermore gather to create loops with each other between layers. Meanwhile, surviving vacancies and interstitials form dislocation loops on the layers. The growth of these inserted layers cause the dimensional increase, i.e. swelling, of graphite. Interstitial and vacancy dislocation loops have been reported and they can easily been observed by electron microscope. However, observation of the intermediate atom clusters becomes is paramount in helping prove this model. We utilize fast protons generated from the University of Missouri Research Reactor (MURR) cyclotron to irradiate highly- oriented pyrolytic graphite (HOPG) as target for this research. Post-irradiation examination (PIE) of dosed targets with high-resolution transmission electron microscopy (HRTEM) has permit observation and analysis of clusters and dislocation loops to support the proposed theory. Another part of the research is to validate M.I. Heggie’s Ruck and Tuck model, which introduced graphite layers may fold under fast particle irradiation. Again, we employed microscopy to image irradiated specimens to determine how the extent of Ruck and Tuck by calculating the number of folds as a function of dose. Our most significant accomplishment is the invention of a novel class of high-intensity pure beta-emitters for long-term lightweight batteries. We have filed four invention disclosure records based on the research conducted in this project. These batteries are lightweight because they consist of carbon and tritium and can be

  7. Studies of mechanical properties and irradiation damage nucleation of HTGR graphites. Progress report, February 1, 1978--March 31, 1979

    International Nuclear Information System (INIS)

    Thrower, P.A.

    1979-04-01

    Irradiation damage effects to highly oriented pyrolytic graphite by 1.5, 5.5 and 15 MeV neutrons are briefly described. Studies of the effects of oxidation on the structure and strength of Stackpole 2020, Great Lakes H440 and POCO AXF-5Q graphites have been continued using H 2 O vapor as the oxidant. The effect of burn-off on compressive strength is similar to that of CO 2 , but structural effects are quite different. The kinetics of the C--CO 2 reaction are being studied using low CO 2 partial pressures in a total pressure of one atmosphere. The necessary modifications to existing apparatus have been made and experimental work is commencing

  8. Graphite Technology Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    W. Windes; T. Burchell; M.Carroll

    2010-10-01

    The Next Generation Nuclear Plant (NGNP) will be a helium-cooled High Temperature Gas Reactor (HTGR) with a large graphite core. Graphite physically contains the fuel and comprises the majority of the core volume. Graphite has been used effectively as a structural and moderator material in both research and commercial high-temperature gas-cooled reactors. This development has resulted in graphite being established as a viable structural material for HTGRs. While the general characteristics necessary for producing nuclear grade graphite are understood, historical “nuclear” grades no longer exist. New grades must be fabricated, characterized, and irradiated to demonstrate that current grades of graphite exhibit acceptable non-irradiated and irradiated properties upon which the thermomechanical design of the structural graphite in NGNP is based. This Technology Development Plan outlines the research and development (R&D) activities and associated rationale necessary to qualify nuclear grade graphite for use within the NGNP reactor.

  9. 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)

  10. HTGR Fuels and Core Development Program. Quarterly progress report for the period ending August 31, 1976. [Graphite and fuel irradiation; fission product release

    Energy Technology Data Exchange (ETDEWEB)

    1976-09-24

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

  11. Temperature and radiolytic corrosion effects on the chlorine behaviour in nuclear graphite: consequences for the disposable of irradiated graphite from UNGG reactors; Effets de la temperature et de la corrosion radiolytique sur le comportement du chlore dans le graphite nucleaire: consequences pour le stockage des graphites irradies des reacteurs UNGG

    Energy Technology Data Exchange (ETDEWEB)

    Vaudey, C.E.

    2010-10-15

    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 ({sup 14}C and {sup 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 {sup 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% {sup 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)

  12. 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

  13. Strength degradation of oxidized graphite support column in VHTR

    International Nuclear Information System (INIS)

    Park, Byung Ha; No, Hee Cheon

    2010-01-01

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

  14. 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

  15. Direct synthesis of graphitic mesoporous carbon from green phenolic resins exposed to subsequent UV and IR laser irradiations

    Science.gov (United States)

    Sopronyi, Mihai; Sima, Felix; Vaulot, Cyril; Delmotte, Luc; Bahouka, Armel; Matei Ghimbeu, Camelia

    2016-12-01

    The design of mesoporous carbon materials with controlled textural and structural features by rapid, cost-effective and eco-friendly means is highly demanded for many fields of applications. We report herein on the fast and tailored synthesis of mesoporous carbon by UV and IR laser assisted irradiations of a solution consisting of green phenolic resins and surfactant agent. By tailoring the UV laser parameters such as energy, pulse repetition rate or exposure time carbon materials with different pore size, architecture and wall thickness were obtained. By increasing irradiation dose, the mesopore size diminishes in the favor of wall thickness while the morphology shifts from worm-like to an ordered hexagonal one. This was related to the intensification of phenolic resin cross-linking which induces the reduction of H-bonding with the template as highlighted by 13C and 1H NMR. In addition, mesoporous carbon with graphitic structure was obtained by IR laser irradiation at room temperature and in very short time periods compared to the classical long thermal treatment at very high temperatures. Therefore, the carbon texture and structure can be tuned only by playing with laser parameters, without extra chemicals, as usually required.

  16. Visible-Light-Irradiated Graphitic Carbon Nitride Photocatalyzed Diels-Alder Reactions with Dioxygen as Sustainable Mediator for Photoinduced Electrons.

    Science.gov (United States)

    Zhao, Yubao; Antonietti, Markus

    2017-08-01

    Photocatalytic Diels-Alder (D-A) reactions with electron rich olefins are realized by graphitic carbon nitride (g-C 3 N 4 ) under visible-light irradiation and aerobic conditions. This heterogeneous photoredox reaction system is highly efficient, and the apparent quantum yield reaches a remarkable value of 47 % for the model reaction. Dioxygen plays a critical role as electron mediator, which is distinct from the previous reports in the homogeneous Ru II complex photoredox system. Moreover, the reaction intermediate vinylcyclobutane is captured and monitored during the reaction, serving as a direct evidence for the proposed reaction mechanism. The cycloaddition process is thereby determined to be the combination of direct [4+2] cycloaddition and [2+2] cycloaddition followed by photocatalytic rearrangement of the vinylcyclobutane intermediate. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  17. WC/Co composite surface structure and nano graphite precipitate induced by high current pulsed electron beam irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Hao, S.Z., E-mail: ebeam@dlut.edu.cn [Key Laboratory of Materials Modification and School of Physics and Optoelectronics Engineering, Dalian University of Technology, Dalian 116024 (China); Zhang, Y. [Key Laboratory of Materials Modification and School of Physics and Optoelectronics Engineering, Dalian University of Technology, Dalian 116024 (China); School of Materials Science and Engineering, Shenyang University of Technology, Shenyang 110870 (China); Xu, Y. [Key Laboratory of Materials Modification and School of Physics and Optoelectronics Engineering, Dalian University of Technology, Dalian 116024 (China); College of Materials Science and Engineering, Liaoning Technical University, Fuxin 123000 (China); Gey, N.; Grosdidier, T. [Université de Lorraine, Laboratoire d’Etude des Microstructures et de Mécanique des Matériaux (LEM3), CNRS UMR 7239, Ile du Saulcy, 57045 Metz (France); Université de Lorraine, Laboratoire d’Excellence on Design of Alloy Metals for Low-Mass Structure (DAMAS), Ile du Saulcy, 57045 Metz (France); Dong, C. [Key Laboratory of Materials Modification and School of Physics and Optoelectronics Engineering, Dalian University of Technology, Dalian 116024 (China); Université de Lorraine, Laboratoire d’Excellence on Design of Alloy Metals for Low-Mass Structure (DAMAS), Ile du Saulcy, 57045 Metz (France)

    2013-11-15

    High current pulsed electron beam (HCPEB) irradiation was conducted on a WC-6% Co hard alloy with accelerating voltage of 27 kV and pulse duration of 2.5 μs. The surface phase structure was examined by using glancing-angle X-ray diffraction (GAXRD), scanning electron microscope (SEM) and high resolution transmission electron microscope (HRTEM) methods. The surface tribological properties were measured. It was found that after 20 pulses of HCPEB irradiation, the surface structure of WC/Co hard alloy was modified dramatically and composed of a mixture of nano-grained WC{sub 1−x}, Co{sub 3}W{sub 9}C{sub 4}, Co{sub 3}W{sub 3}C phases and graphite precipitate domains ∼50 nm. The friction coefficient of modified surface decreased to ∼0.38 from 0.6 of the initial state, and the wear rate reduced from 8.4 × 10{sup −5} mm{sup 3}/min to 6.3 × 10{sup −6} mm{sup 3}/min, showing a significant self-lubricating effect.

  18. Toxicological characterization of chemicals produced from laser irradiation of graphite composite materials

    International Nuclear Information System (INIS)

    Kwan, J.

    1990-11-01

    One of the major potential hazards associated with laser machining of graphite composite materials is the toxic fumes and gases that are generated. When exposed to the intense energy of the laser beam, the organic polymer matrix of the composite material may decompose into various toxic by-products. To advance the understanding of the laser machining process from a health and safety viewpoint, this particular study consisted of the following steps: collect and analyze gaseous by-products generated during laser machining; collect particulates generated during laser machining and chemically extract them to determine the chemical species that may have absorbed or recondensed onto these particles; and review and evaluate the toxicity of the identified chemical species

  19. 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)

  20. Fabrication of SnO₂-reduced graphite oxide monolayer-ordered porous film gas sensor with tunable sensitivity through ultra-violet light irradiation.

    Science.gov (United States)

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

    2015-03-11

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

  1. Experimental Plan for EDF Energy Creep Rabbit Graphite Irradiations- Rev. 2 (replaces Rev. 0 ORNL/TM/2013/49).

    Energy Technology Data Exchange (ETDEWEB)

    Burchell, Timothy D [ORNL

    2014-07-01

    The experimental results obtained here will assist in the development and validation of future models of irradiation induced creep of graphite by providing the following data: Inert creep stain data from low to lifetime AGR fluence Inert creep-property data (especially CTE) from low to lifetime AGR fluence Effect of oxidation on creep modulus (by indirect comparison with experiment 1 and direct comparison with experiment 3 NB. Experiment 1 and 3 are not covered here) Data to develop a mechanistic understanding, including oAppropriate creep modulus (including pinning and high dose effects on structure) oInvestigation of CTE-creep strain behavior under inert conditions oInformation on the effect of applied stress/creep strain on crystallite orientation (requires XRD) oEffect of creep strain on micro-porosity (requires tomography & microscopy) This document describes the experimental work planned to meet the requirements of project technical specification [1] and EDF Energy requests for additional Pre-IE work. The PIE work is described in detail in this revision (Section 8 and 9).

  2. Graphite targets at LAMPF

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  3. 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

  4. Organic matters removal from landfill leachate by immobilized Phanerochaete chrysosporium loaded with graphitic carbon nitride under visible light irradiation.

    Science.gov (United States)

    Hu, Liang; Liu, Yutang; Zeng, Guangming; Chen, Guiqiu; Wan, Jia; Zeng, Yunxiong; Wang, Longlu; Wu, Haipeng; Xu, Piao; Zhang, Chen; Cheng, Min; Hu, Tianjue

    2017-10-01

    This study investigated the technical applicability of a combination of Phanerochaete chrysosporium (P. chrysosporium) with photocatalyst graphitic carbon nitride (g-C 3 N 4 ) for organic matters removal from landfill leachate under visible light irradiation. Photocatalyst g-C 3 N 4 was well immobilized on the hyphae surface of P. chrysosporium by calcium alginate. The typical absorption edge in visible light region for g-C 3 N 4 was at about 460 nm, and the optical absorption bandgap of g-C 3 N 4 was estimated to be 2.70 eV, demonstrating the great photoresponsive ability of g-C 3 N 4 . An optimized g-C 3 N 4 content of 0.10 g in immobilized P. chrysosporium and an optimized immobilized P. chrysosporium dosage of 1.0 g were suitable for organic matters removal. The removal efficiency of total organic carbon (TOC) reached 74.99% in 72 h with the initial TOC concentration of 100 mg L -1 . In addition, the gas chromatography coupled with mass spectrometry (GC-MS) measurements showed that immobilized P. chrysosporium presented an outstanding removal performance for almost all organic compounds in landfill leachate, especially for the volatile fatty acids and long-chain hydrocarbons. The overall results indicate that the combination P. chrysosporium with photocatalyst g-C 3 N 4 for organic matters removal from landfill leachate may provide a more comprehensive potential for the landfill leachate treatment. Copyright © 2017 Elsevier Ltd. All rights reserved.

  5. Graphite moderator annealing of the experimental reactor for irradiation (0.5 MW); Recozimento do moderador de grafite do reator experimental de irradiacao (0.5MW)

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira Avila, Carlos Alberto de; Pires, Luis Fernando Goncalves [Centro Tecnologico do Exercito, Rio de Janeiro, RJ (Brazil). Instituto de Projetos Especiais

    1995-12-31

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

  6. 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)

  7. Evaluation of plasma disruption simulating short pulse laser irradiation experiments on boronated graphites and CFCs [carbon fibre composites

    International Nuclear Information System (INIS)

    Stad, R.C.L. van der; Klippel, H.T.; Kraaij, G.J.

    1992-12-01

    New experimental and numerical results from disruption heat flux simulations in the millisecond range with laser beams are discussed. For a number of graphites, boronated graphites and carbon fibre composites, the effective enthalpy of ablation is determined as 30 ± 3 MJ/kg, using laser pulses of about -.3 ms. The numerical results predict the experimental results rather well. No effect of boron doping on the ablation enthalpy is found. (author). 9 refs., 4 figs., 1 tab

  8. Graphite oxidation and structural strength of bottom support system in VHTR

    International Nuclear Information System (INIS)

    Park, Byung Ha

    2009-02-01

    The air-ingress event by a large pipe break is an important accident considered in design of very high-temperature gas-cooled reactors (VHTR). Core collapse prediction is a main safety issue. Graphite oxidation model and structural failure model are technically required. In this study, one main target is to develop graphite oxidation model based on Arrhenius model for graphite oxidation reaction. Kinetic parameters of IG-430 were measured. Temperature was controlled at the range of 540 to 800 .deg. C to estimate activation energy. Oxygen concentration was controlled over the range of 0 to 0.34 mole fraction to find order of reaction. It turns out that activation energy Ea was 258.15 ± 1.5 kJ/mol and order of reaction was 0.37 ± 0.04. The other main target is to develop the structural failure model. Graphite support column is important for VHTR structural integrity. Graphite support columns are under the axial load. Critical strength of graphite column is related to slenderness ratio and bulk density. Through compression tests for fresh and oxidized graphite columns we show that compressive strength of IG-110 was 79.46 MPa. And, the buckling strength of IG-110 column was expressed by the empirical formula. σ 0 = σ straight-line -C r L σ straight-live = 91.31 MPa, C=1.01 The results of uniform and non-uniform oxidation tests show that the strength degradation of oxidized graphite column is expressed in the following non-dimensional form: σ 0 /σ = exp(-kd), k=0.111 Also, from the results of the uniform oxidation test with a complicated-shape column, we found out that the above non-dimensional equation obtained from the uniform oxidation test is applicable to a uniform oxidation case with a complicated-shape column

  9. Graphites for nuclear applications; Les graphites pour les applications nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Bonal, J.P.; Gosmain, L. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DMN), Lab. de Microscopie et d' Etudes de l' Endommagement, 91 - Gif-sur-Yvette (France)

    2006-03-15

    Being an excellent neutron moderator, graphite is used as a structural material in many nuclear reactor types. By the end of the 50's, the French gas-cooled reactor development needed manufacturing of a nuclear-grade graphite. Graphite irradiation can lead to in-lattice energy accumulation, dimensional changes and physical properties modification. Moreover, the radiolytic corrosion induced by the coolant (CO{sub 2}) may generate mechanical properties degradation. Today, French gas-cooled reactors are all in their decommissioning phase that requires the knowledge of the radiological inventory of the irradiated graphites. At present time, graphite is still foreseen as a future material for hydrogen production by high temperature gas cooled nuclear plants. In the future, graphite will be the necessary moderator material for high temperature reactors with thermal neutron spectrum dedicated to hydrogen and electricity production. (authors)

  10. AGC-2 Graphite Preirradiation Data Package

    Energy Technology Data Exchange (ETDEWEB)

    David Swank; Joseph Lord; David Rohrbaugh; William Windes

    2012-10-01

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

  11. Graphite surveillance in N Reactor

    International Nuclear Information System (INIS)

    Woodruff, E.M.

    1991-09-01

    Graphite dimensional changes in N Reactor during its 24 yr operating history are reviewed. Test irradiation results, block measurements, stack profiles, top of reflector motion monitors, and visual observations of distortion are described. 18 refs., 14 figs., 1 tab

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

    International Nuclear Information System (INIS)

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

    1987-04-01

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  14. 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)

  15. AGC-2 Irradiation Report

    Energy Technology Data Exchange (ETDEWEB)

    Rohrbaugh, David Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Windes, William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-06-01

    The Next Generation Nuclear Plant (NGNP) will be a helium-cooled, very high temperature reactor (VHTR) with a large graphite core. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs.[ , ] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphites have been developed and are considered suitable candidates for the new NGNP reactor design. To support the design and licensing of NGNP 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 related to the life limiting effects of irradiation creep on key physical properties of the NGNP candidate graphites. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change. In this case, the graphite physically changes i.e., first shrinking and then expanding with increasing neutron dose. This disparity in material volume change can induce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the

  16. 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

  17. 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

  18. Study by electronic microscopy of corrosion features of graphite after hot oxidation (air, 620 C)

    International Nuclear Information System (INIS)

    Jodon de Villeroche, Suzanne

    1968-01-01

    The author reports the study of corrosion features of graphite after hot oxidation in the air at 620 C. It is based on observations made by electronic microscopy. This study comes after another one dedicated to oxidation features obtained by hot corrosion of natural graphite, and aims at comparing pyrolytic graphite before and after irradiation in an atomic pile, and at performing tests on a graphite processed with ozone. After a recall of generalities about natural graphite and of some issues related to hot corrosion of natural graphite, the author presents some characteristics and features of irradiated and non-irradiated pyrolytic graphite. He reports the study of the oxidation of samples of pyrolytic graphite: production of thin lamellae, production of glaze-carbon replicates, oxidation of irradiated and of non-irradiated graphite, healing of irradiation defects, and oxidation of ozone-processed natural graphite [fr

  19. Ultrasonic irradiation preparation of graphitic-C3N4/polyaniline nanocomposites as counter electrodes for dye-sensitized solar cells.

    Science.gov (United States)

    Afshari, Mohaddeseh; Dinari, Mohammad; Momeni, Mohamad Mohsen

    2018-04-01

    In this research, polyaniline/graphitic carbon nitride (PANI/g-C 3 N 4 ) nanocomposites were synthesized via in-situ electrochemical polymerization of aniline monomer whit different number of cyclic voltammetry scans (10, 20 and 30 cycles) after electrode surface pre-preparation using a potential shock under ultrasonic irradiation. PANI/g-C 3 N 4 nanocomposites with two values of g-C 3 N 4 (0.010 wt% and 0.015 wt%) were deposited on the surface of the transparent conducting film (FTO glass) by immersing FTO into the aniline solution and g-C 3 N 4 during the electro-polymerization. The resulting PANI/g-C 3 N 4 films were characterized by Fourier transformed infra-red (FTIR), power X-ray diffraction (PXRD), field emission scanning electron microscopy (FE-SEM) and transmission electron microscopy (TEM) techniques. The prepared electrodes were applied as counter electrode in dye-sensitized solar cells. Among them, the prepared electrode with 10 cycles and 0.01 wt% g-C 3 N 4 showed the best efficiency. These hybrids show good catalytic activity in elevating tri-iodide reduction and due to the synergistic effect of PANI and g-C 3 N 4 , PANI/g-C 3 N 4 nanocomposite electrode shows power conversion efficiency about 1.8%. Copyright © 2017 Elsevier B.V. All rights reserved.

  20. Graphitic carbon nitride (g-C3N4)-Pt-TiO2 nanocomposite as an efficient photocatalyst for hydrogen production under visible light irradiation.

    Science.gov (United States)

    Chai, Bo; Peng, Tianyou; Mao, Jing; Li, Kan; Zan, Ling

    2012-12-28

    Porous graphitic carbon nitride (g-C(3)N(4)) was prepared by a simple pyrolysis of urea, and then a g-C(3)N(4)-Pt-TiO(2) nanocomposite was fabricated via a facile chemical adsorption followed by a calcination process. The obtained products were characterized by X-ray diffraction, X-ray photoelectron spectroscopy, UV-vis diffuse reflectance absorption spectra, and electron microscopy. It is found that the visible-light-induced photocatalytic hydrogen evolution rate can be remarkably enhanced by coupling TiO(2) with the above g-C(3)N(4), and the g-C(3)N(4)-Pt-TiO(2) composite with a mass ratio of 70 : 30 has the maximum photoactivity and excellent photostability for hydrogen production under visible-light irradiation, and the stable photocurrent of g-C(3)N(4)-TiO(2) is about 1.5 times higher than that of the bare g-C(3)N(4). The above experimental results show that the photogenerated electrons of g-C(3)N(4) can directionally migrate to Pt-TiO(2) due to the close interfacial connections and the synergistic effect existing between Pt-TiO(2) and g-C(3)N(4) where photogenerated electrons and holes are efficiently separated in space, which is beneficial for retarding the charge recombination and improving the photoactivity.

  1. 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.

  2. Sugarcane juice derived carbon dot–graphitic carbon nitride composites for bisphenol A degradation under sunlight irradiation

    Science.gov (United States)

    Wong, Jing Lin; Hak, Chen Hong; Tai, Jun Yan; Leong, Kah Hon; Saravanan, Pichiah

    2018-01-01

    Carbon dots (CDs) and graphitic carbon nitride (g-C3N4) composites (CD/g-C3N4) were successfully synthesized by a hydrothermal method using urea and sugarcane juice as starting materials. The chemical composition, morphological structure and optical properties of the composites and CDs were characterized using various spectroscopic techniques as well as transmission electron microscopy. X-ray photoelectron spectroscopy (XPS) results revealed new signals for carbonyl and carboxyl groups originating from the CDs in CD/g-C3N4 composites while X-ray diffraction (XRD) results showed distortion of the host matrix after incorporating CDs into g-C3N4. Both analyses signified the interaction between g-C3N4 and CDs. The photoluminescence (PL) analysis indicated that the presence of too many CDs will create trap states at the CD/g-C3N4 interface, decelerating the electron (e−) transport. However, the CD/g-C3N4(0.5) composite with the highest coverage of CDs still achieved the best bisphenol A (BPA) degradation rate at 3.87 times higher than that of g-C3N4. Hence, the charge separation efficiency should not be one of the main factors responsible for the enhancement of the photocatalytic activity of CD/g-C3N4. Instead, the light absorption capability was the dominant factor since the photoreactivity correlated well with the ultraviolet–visible diffuse reflectance spectra (UV–vis DRS) results. Although the CDs did not display upconversion photoluminescence (UCPL) properties, the π-conjugated CDs served as a photosensitizer (like organic dyes) to sensitize g-C3N4 and injected electrons to the conduction band (CB) of g-C3N4, resulting in the extended absorption spectrum from the visible to the near-infrared (NIR) region. This extended spectral absorption allows for the generation of more electrons for the enhancement of BPA degradation. It was determined that the reactive radical species responsible for the photocatalytic activity were the superoxide anion radical (O2

  3. Sugarcane juice derived carbon dot-graphitic carbon nitride composites for bisphenol A degradation under sunlight irradiation.

    Science.gov (United States)

    Sim, Lan Ching; Wong, Jing Lin; Hak, Chen Hong; Tai, Jun Yan; Leong, Kah Hon; Saravanan, Pichiah

    2018-01-01

    Carbon dots (CDs) and graphitic carbon nitride (g-C 3 N 4 ) composites (CD/g-C 3 N 4 ) were successfully synthesized by a hydrothermal method using urea and sugarcane juice as starting materials. The chemical composition, morphological structure and optical properties of the composites and CDs were characterized using various spectroscopic techniques as well as transmission electron microscopy. X-ray photoelectron spectroscopy (XPS) results revealed new signals for carbonyl and carboxyl groups originating from the CDs in CD/g-C 3 N 4 composites while X-ray diffraction (XRD) results showed distortion of the host matrix after incorporating CDs into g-C 3 N 4 . Both analyses signified the interaction between g-C 3 N 4 and CDs. The photoluminescence (PL) analysis indicated that the presence of too many CDs will create trap states at the CD/g-C 3 N 4 interface, decelerating the electron (e - ) transport. However, the CD/g-C 3 N 4 (0.5) composite with the highest coverage of CDs still achieved the best bisphenol A (BPA) degradation rate at 3.87 times higher than that of g-C 3 N 4 . Hence, the charge separation efficiency should not be one of the main factors responsible for the enhancement of the photocatalytic activity of CD/g-C 3 N 4 . Instead, the light absorption capability was the dominant factor since the photoreactivity correlated well with the ultraviolet-visible diffuse reflectance spectra (UV-vis DRS) results. Although the CDs did not display upconversion photoluminescence (UCPL) properties, the π-conjugated CDs served as a photosensitizer (like organic dyes) to sensitize g-C 3 N 4 and injected electrons to the conduction band (CB) of g-C 3 N 4 , resulting in the extended absorption spectrum from the visible to the near-infrared (NIR) region. This extended spectral absorption allows for the generation of more electrons for the enhancement of BPA degradation. It was determined that the reactive radical species responsible for the photocatalytic activity were

  4. 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

  5. Mechanical properties of graphite and carbon materials

    International Nuclear Information System (INIS)

    Jouquet, G.

    1976-01-01

    The elastic properties of the graphite monocrystal, the role of internal characteristics (texture, porosity) on the mechanical behavior of carbons, effects caused by the gaseous environment and neutron irradiation, and the resistance of graphites to cyclic mechanical stresses are discussed [fr

  6. Compressive loading unloading behavior of nuclear graphite grades of different forming method and raw cokes

    International Nuclear Information System (INIS)

    Chi, Sehwan; Hong, Seongdeok; Kim, Yongwan

    2012-01-01

    Nuclear graphite is used for core structural components and neutron moderators in high temperature gas-cooled reactors. As graphite is a brittle material fail at relatively low strains (e.g., ∼0.5% in tension and ∼2% in compression), cracking of these components can occur throughout the life of the reactor under the influence of thermal and mechanical stresses. While a lot of studies have been performed on the fracture of graphite, most studies have been concerned on crack initiation and propagation, with little concerns on the damage processes that lead to the very first stage of crack initiation. In this study, the graphite damage processes before the main crack formation were investigated based on the microstructure change during load relaxation. For this, 4-1/3 notched flexure strength test specimens made of nuclear graphite grades IG-110, NBG-18 and PCEA of different forming methods (isotropic molding, vibrational molding and extrusion, respectively) and ingredients (coke, binder) were subjected to 10 cyclic compressive loading-unloading, and the changes in the microstructure of notch-tip areas were examined by X-ray tomography

  7. 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.

  8. 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.

  9. Experience with graphite in JET

    International Nuclear Information System (INIS)

    Pick, M.A.; Celentano, G.; Deksnis, E.; Dietz, K.J.; Shaw, R.; Sonnenberg, K.; Walravens, M.

    1987-01-01

    During the current operational period of JET more than 50% of the internal area of the machine is covered in graphite tiles. This includes the 15 m 2 of carbon tiles installed in the new toroidal limiter, the 40 poloidal belts of graphite tiles covering the U-joints and bellows as well as a two metre high ring (-- 20 m 2 ) or carbon tiles on the inner wall of the Torus. A ring of tiles in the equatorial plane (3 tiles high) consists of carbon-carbon fibre tiles. Test bed results indicated that the fine grained graphite tiles cracked at ∼ 1 kW/cm 2 for 2s of irradiation whereas the carbon-carbon fibre tiles were able to sustain a flux, limited by the irradiation facility, of 3.5 kW for 3s without any damage. The authors report on the generally positive experience they have had had with the installed graphite during the present and previous in-vessel configurations. This includes the physical integrity of the tiles under severe conditions such as high energy run-away electron beams, plasma disruptions and high heat fluxes. They report on the importance of the precise positioning of the inner wall and x-point tiles at the very high power fluxes of JET and the effect of deviations on both graphite and carbon-fibre tiles

  10. Progress in radioactive graphite waste management

    International Nuclear Information System (INIS)

    2010-07-01

    , especially in the UK. It is intended that this report which contains the proceedings of the conference should contribute to progress in the management of radioactive graphite worldwide. The report contains a selection of the papers presented on various issues related to dismantling and treating irradiated graphite. In addition, the report contains summaries of the four topical discussions which were held during the conference

  11. 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}.

  12. NGNP Graphite Selection and Acquisition Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Burchell, T.; Bratton, R.; Windes, W.

    2007-09-30

    The nuclear graphite (H-451) previously used in the United States for High-Temperature Reactors (HTRs) is no longer available. New graphites have been developed and are considered suitable candidates for the Next-Generation Nuclear Plant (NGNP). A complete properties database for these new, available, candidate grades of graphite must be developed to support the design and licensing of NGNP core components. Data are required for the physical, mechanical (including radiation-induced creep), and oxidation properties of graphites. Moreover, the data must be statistically sound and take account of in-billet, between billets, and lot-to-lot variations of properties. These data are needed to support the ongoing development1 of the risk-derived American Society of Mechanical Engineers (ASME) graphite design code (a consensus code being prepared under the jurisdiction of the ASME by gas-cooled reactor and NGNP stakeholders including the vendors). The earlier Fort St. Vrain design of High-Temperature Reactor (HTRs) used deterministic performance models for H-451, while the NGNP will use new graphite grades and risk-derived (probabilistic) performance models and design codes, such as that being developed by the ASME. A radiation effects database must be developed for the currently available graphite materials, and this requires a substantial graphite irradiation program. The graphite Technology Development Plan (TDP)2 describes the data needed and the experiments planned to acquire these data in a timely fashion to support NGNP design, construction, and licensing. The strategy for the selection of appropriate grades of graphite for the NGNP is discussed here. The final selection of graphite grades depends upon the chosen reactor type and vendor because the reactor type (pebble bed or prismatic block) has a major influence on the graphite chosen by the designer. However, the time required to obtain the needed irradiation data for the selected NGNP graphite is sufficiently

  13. 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

  14. 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

  15. Radiation behaviour of graphite for HTGR

    International Nuclear Information System (INIS)

    Shtrombakh, Ya.I.; Platonov, P.A.; Gurovich, B.A.; Alekseev, V.M.

    1996-01-01

    The paper presents the results of investigations of different graphite materials, among with the standard reactor graphite manufacturing by electrode technology and a number of advanced graphites of new generation. During the investigation of radiation stability of standard reactor graphite the basic mechanisms of radiation damage of its structure were studied. With the help of transmission electron microscopy deformations and cracking of filler and binder were detected in the vicinity of the boundaries, separating these two components. Cracking begins with crystallite splitting and ends in full fracture of boundary layers. Such process of degradation can be explained by disjoint deformations resulting from difference in growth rate of filler and binder crystallites, in its turn caused by considerable difference between their sizes. It has been concluded that radiation stability of graphite may be improved by creating such graphite materials, in which the difference in sizes of crystallites of different structure components would be the minimal possible. When developing production technology of isotropic graphite for high temperature reactors, some progress was made towards the solution of this problem. Despite considerable swelling at high temperature this type of graphite appeared to be substantially less susceptible to the degradation of the structure and to deterioration of physico-mechanical properties. In addition to graphites manufactured by tradition technology, the graphite was investigated, in which pyrocarbon precipitated from gas phase under 1000 deg. C was used as binder. Carbon precipitated in such a way was non-graphitized at high temperatures and therefore it demonstrated sharp shrinkage under irradiation at high temperature, and shrinkage rate correlated with pyrocarbon quota in graphite structure. (author). 5 refs, 18 figs, 1 tab

  16. 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.

  17. 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

  18. The effect of volume on the tensile strength of several nuclear-grade graphites

    International Nuclear Information System (INIS)

    Strizak, J.P.

    1991-01-01

    This report will present the results of a study on the effects of stress volume on the tensile strength of two nuclear-grade graphites. The materials selected were H-451, an extruded near-isotropic graphite manufactured by Great Lakes Carbon Corporation, and IG-110, a fine-grained isotropic molded graphite manufactured by Toyo Tanso Company, Ltd. The tensile properties of H-451 were examined extensively in the past in order to characterize the variability of strength within billets, between billets, and between lots. But, the variability within a billet was, for the most part, studied only casually. The problem was the strong influence of a limited sampling plan in describing the mean strength and the variability. Therefore, an extensive, statistically sound sampling plan has been devised to fully characterize the spatial variability within a single billet. The effects of stress volume are being reexamined by comparing the strengths of four specimen sizes covering a broad range in stress volume. Two models will be employed for analysis of the stress volume data for the selected graphites. The popular Weibull model has previously been found to grossly overestimate the volume dependence of the strength of H-451 graphite. The model will be reevaluated using the improved statistical distribution of strength expected from the current sampling plan. A new fracture model developed by Burchell and Tucker has potential for determining the effect of stress volume on the tensile strength of graphite. This probabilistic failure criterion combines a microstructural basis with a fracture-mechanics approach to failure. An initial evaluation of H-451 data showed that the model closely predicted the mean tensile strength for the two smaller specimen sizes. 9 refs., 24 figs., 1 tab

  19. Graphite target for the spiral project

    Energy Technology Data Exchange (ETDEWEB)

    Putaux, J.C.; Ducourtieux, M.; Ferro, A.; Foury, P.; Kotfila, L.; Mueller, A.C.; Obert, J.; Pauwels, N.; Potier, J.C.; Proust, J. [Paris-11 Univ., 91 - Orsay (France). Inst. de Physique Nucleaire; Bertrand, P. [Grand Accelerateur National d`Ions Lourds (GANIL), 14 - Caen (France); Loiselet, M. [Universite Catholique de Louvain, Louvain-La-Neuve (Belgium)] [and others

    1996-12-31

    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). 5 refs.

  20. Graphite reactor physics

    International Nuclear Information System (INIS)

    Bacher, P.; Cogne, F.

    1964-01-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  2. Theoretical basis for graphite stress analysis in BERSAFE

    International Nuclear Information System (INIS)

    Harper, P.G.

    1980-03-01

    The BERSAFE finite element computer program for structural analysis has been extended to deal with structures made from irradiated graphite. This report describes the material behaviour which has been modelled and gives the theoretical basis for the solution procedure. (author)

  3. Analysis of Picosecond Pulsed Laser Melted Graphite

    Science.gov (United States)

    Steinbeck, J.; Braunstein, G.; Speck, J.; Dresselhaus, M. S.; Huang, C. Y.; Malvezzi, A. M.; Bloembergen, N.

    1986-12-01

    A Raman microprobe and high resolution TEM have been used to analyze the resolidified region of liquid carbon generated by picosecond pulse laser radiation. From the relative intensities of the zone center Raman-allowed mode for graphite at 1582 cm{sup -1} and the disorder-induced mode at 1360 cm{sup -1}, the average graphite crystallite size in the resolidified region is determined as a function of position. By comparison with Rutherford backscattering spectra and Raman spectra from nanosecond pulsed laser melting experiments, the disorder depth for picosecond pulsed laser melted graphite is determined as a function of irradiating energy density. Comparisons of TEM micrographs for nanosecond and picosecond pulsed laser melting experiments show that the structure of the laser disordered regions in graphite are similar and exhibit similar behavior with increasing laser pulse fluence.

  4. Analysis of picosecond pulsed laser melted graphite

    Energy Technology Data Exchange (ETDEWEB)

    Steinbeck, J.; Braunstein, G.; Speck, J.; Dresselhaus, M.S.; Huang, C.Y.; Malvezzi, A.M.; Bloembergen, N.

    1986-01-01

    A Raman microprobe and high resolution TEM have been used to analyze the resolidified region of liquid carbon generated by picosecond pulse laser radiation. From the relative intensities of the zone center Raman-allowed mode for graphite at 1582 cm/sup -1/ and the disorder-induced mode at 1360 cm/sup -1/, the average graphite crystallite size in the resolidified region is determined as a function of position. By comparison with Rutherford backscattering spectra and Raman spectra from nonosecond pulsed laser melting experiments, the disorder depth for picosecond pulsed laser melted graphite is determined as a function of irradiating energy density. Comparisons of TEM micrographs for nanosecond and picosecond pulsed laser melting experiments show that the structure of the laser disordered regions in graphite are similar and exhibit similar behavior with increasing laser pulse fluence.

  5. Analysis of picosecond pulsed laser melted graphite

    International Nuclear Information System (INIS)

    Steinbeck, J.; Braunstein, G.; Speck, J.; Dresselhaus, M.S.; Huang, C.Y.; Malvezzi, A.M.; Bloembergen, N.

    1986-01-01

    A Raman microprobe and high resolution TEM have been used to analyze the resolidified region of liquid carbon generated by picosecond pulse laser radiation. From the relative intensities of the zone center Raman-allowed mode for graphite at 1582 cm -1 and the disorder-induced mode at 1360 cm -1 , the average graphite crystallite size in the resolidified region is determined as a function of position. By comparison with Rutherford backscattering spectra and Raman spectra from nonosecond pulsed laser melting experiments, the disorder depth for picosecond pulsed laser melted graphite is determined as a function of irradiating energy density. Comparisons of TEM micrographs for nanosecond and picosecond pulsed laser melting experiments show that the structure of the laser disordered regions in graphite are similar and exhibit similar behavior with increasing laser pulse fluence

  6. Studies on mechanical properties of graphites for HTGR

    International Nuclear Information System (INIS)

    Oku, T.; Eto, M.; Fujisaki, K.; Yoda, S.; Ishiyama, S.; Sugihara, T.

    1982-01-01

    Recent research on the mechanical properties of HTGR graphites at JAERI is reviewed. The mechanical properties of graphites are required for predicting the stresses induced in the core graphite structures during reactor operation and for evaluating non-failure probabilities of the graphite structures. In this paper, effects of irradiation, stress and oxidation on the mechanical properties and fatigue properties of petroleum coke semi-isotropic and isotropic graphites for HTGRs are primarly described. Young's modulus and bend strength before and after neutron irradiation have been measured to examine irradiation effects on a fracture criterion of graphites. Two kinds of relationships are found between the bend strength and Young's modulus, depending upon the irradiation temperature. Changes in Young's modulus after irradiation are found to be different from those after irradiation creep deformation. Young's modulus under compressive stress is equivalent to that at the onset of unloading. Oxidation gives rise to the decreases in density and modulus, and also brings about a strength degradation. Tension-compression fatigue strengths are obtained and arranged successfully using statistical trivariant method with tension-tension fatigue strength data

  7. 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.

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

    International Nuclear Information System (INIS)

    Elhardt, W.

    1979-01-01

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

  9. 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

  10. Progress in radioactive graphite waste management. Additional information

    International Nuclear Information System (INIS)

    2010-06-01

    , especially in the UK. It is intended that this report which contains the proceedings of the conference should contribute to progress in the management of radioactive graphite worldwide. The report contains a selection of the papers presented on various issues related to dismantling and treating irradiated graphite. In addition, the report contains summaries of the four topical discussions which were held during the conference

  11. Diamond amorphization in neutron irradiation

    International Nuclear Information System (INIS)

    Nikolaenko, V.A.; Gordeev, V.G.

    1996-01-01

    The paper presents the results on neutron irradiation of the diamond in a nuclear reactor. It is shown that the neutron irradiation stimulates the diamond transition to the amorphous state. At a temperature below 750 o K the time required for the diamond-graphite transition decreases with decreasing irradiation temperature. On the contrary, in irradiation at higher temperatures the time of diamond conversion into the amorphous state increases with decreasing but always remains shorter than in the absence of irradiation. (author)

  12. Characterization of graphite dust produced by pneumatic lift

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

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

  13. Topological investigation of nuclear graphite using small angle scattering

    Science.gov (United States)

    Rai, Durgesh K.; Khaykovich, Boris; Campbell, Anne A.; Ilvasky, Jan; Katoh, Yutai; Snead, Lance L.

    Nuclear power reactors require high performance materials that withstand high temperatures and neutron damage over long period of times. Graphite is widely used for high temperature fission reactor applications. It has a complex multiphase microstructure, which is affected by neutron irradiation. The irradiation-induced microstructures result in significant thermophysical property changes, affecting service lifetimes. It is important to understand these life-limiting phenomena at many different length scales. We present the results from small angle scattering (SAS) studies on graphite samples, which vary in doses and irradiation temperatures. The neutron and synchrotron SAS measurement data indicates that the graphite morphology consists of surface fractal structures. The samples were found to be uniform across several decades of length scale, while exhibiting different surface fractal dimensions, for different irradiation doses and temperature conditions. The surface fractal dimension changes at HFIR at ORNL, DOE User Facility; APS at ANL, DOE User Facility; Office of Nuclear Energy NSUF.

  14. 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.

  15. Facile one-pot synthesis of cerium oxide/sulfur-doped graphitic carbon nitride (g-C3N4) as efficient nanophotocatalysts under visible light irradiation.

    Science.gov (United States)

    Jourshabani, Milad; Shariatinia, Zahra; Badiei, Alireza

    2017-12-01

    Porous CeO 2 /sulfur-doped g-C 3 N 4 (CeO 2 /CNS) composites were synthesized by one-pot thermal condensation of thiourea and cerium nitrate as starting materials. The obtained CeO 2 (x)/CNS composites (x=8.4, 9.5 and 10.4wt%) with different CeO 2 contents were characterized by the XRD, FT-IR, XPS, TEM, BET, DRS and PL analyses. The TEM images displayed a nonporous and platelet-like morphology for pure CNS but a nanoporous structure with numerous uniform pore sizes of ∼40nm for the CeO 2 (9.5)/CNS composite. The XRD phase structures and TEM morphologies confirmed that structural evolution trend and stacking degree of CNS were disrupted in precense of the CeO 2 nanoparticles. The optimized photocatalyst, i.e. CeO 2 (9.5)/CNS nanocomposite, exhibited the highest visible light photocatalytic activity (91.4% after 150min) with a reaction rate constant of 0.0152min -1 toward methylene blue (MB) degradation which was greater compared with the individual CNS (0.0044min -1 ) and CeO 2 (0.0031min -1 ) photocatalysts. This enhanced photocatalytic performance was originated from heterojunctions formed between CeO 2 and CNS that improved the effective charge transfer through interfacial interactions between both components. The heterojunction prepared displayed excellent stability for the photocatalytic activity under the optimized conditions including catalyst dosage 0.08g, initial dye concentration 7mg/L and irradiation time 150min which was obtained using response surface methodology (RSM). The trapping experiments using isopropanol, benzoquinone and ethylenediaminetetraacetic as the OH, O 2 - and h + scavengers, respectively, verified that the OH and O 2 - as major species directly attacked onto the MB molecules while h + showed a negligible role. Finally, it could be stated that simultaneous doping of both sulfur and CeO 2 within the g-C 3 N 4 structure using a simple one-pot synthetic process produced very active photocatalysts illustrating their potential for

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

    Science.gov (United States)

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

    2012-05-08

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

  17. 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...

  18. Graphitic Carbon Nanocubes Derived from ZIF-8 for Photothermal Therapy.

    Science.gov (United States)

    Chen, Wei; Zhang, Xiaoman; Ai, Fujin; Yang, Xueqing; Zhu, Guangyu; Wang, Feng

    2016-06-20

    Graphitic carbon nanocubes (GCNCs) were prepared by pyrolysis of ZIF-8 nanocubes. The GCNCs resemble the structure of N-doped graphite and exhibit a high photothermal conversion efficiency of 40.4%. In vitro tests demonstrate that the GCNCs are highly biocompatible and induce an effective photothermal therapy effect under 808 nm irradiation. Our study provides a facile strategy for preparing functional carbon nanomaterials of prescribed size, morphology, and porous structure for bioapplications.

  19. Electrochemical treatment of graphite

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  20. Assessment of management modes for graphite from reactor decommissioning

    International Nuclear Information System (INIS)

    White, I.F.; Smith, G.M.; Saunders, L.J.; Kaye, C.J.; Martin, T.J.; Clarke, G.H.; Wakerley, M.W.

    1984-01-01

    A technological and radiological assessment has been made of the management options for irradiated graphite wastes from the decommissioning of Magnox and advanced gas-cooled reactors. Detailed radionuclide inventories have been estimated, the main contribution being from activation of the graphite and its stable impurities. Three different packaging methods for graphite have been described; each could be used for either sea or land disposal, is logistically feasible and could be achieved at reasonable cost. Leaching tests have been carried out on small samples of irradiated graphite under a variety of conditions including those of the deep ocean bed; the different conditions had little effect on the observed leach rates of radiologically significant radionuclides. Radiological assessments were made of four generic options for disposal of packaged graphite: on the deep ocean bed, in deep geologic repositories at two different types of site, and by shallow land burial. Incineration of graphite was also considered, though this option presents logistical problems. With appropriate precautions during the lifetime of the Cobalt-60 content of the graphite, any of the options considered could give acceptably low doses to individuals, and all would merit further investigation in site-specific contexts

  1. TSX graphite for extended use in the N-Reactor

    International Nuclear Information System (INIS)

    Kennedy, C.R.

    1985-08-01

    This report reviews the limited amount of irradiation data available for grade TSX graphite with the purpose of obtaining reasonable estimates of material behavior. The results are enhanced by obtaining generalized behavior characteristics demonstrated by similar grades of graphite, such as CSF, AGOT, and PGA. Intent of this work is to furnish the necessary coefficients to describe the material behavior for inclusion in the constitutive equations for the anisotropic graphite grade TSX. Estimates of the free-dimensional changes of TSX graphite as a function of temperature and fluence have been made and shown to be in good agreement with the data. The effects of irradiation on other physical properties, such as elastic moduli, conductivity, and coefficient of thermal expansion, are also described. The irradiation creep characteristics of TSX graphite are also estimated on the basis of data for similar grades of graphite in the US and Europe. Crude approximations of stresses generated in the keyed structure were made to demonstrate the magnitude of the problem. The results clearly predict that the filler-block keys will fail and the tube-block keys will not. It is also indicated that the overall stack height growth will be increased by 25 to 38 mm (1 to 1.5 in.) because of creep

  2. Oxidizability and explosibility of pure graphite powder

    International Nuclear Information System (INIS)

    L Rahmani; D Roubineau; S Cornet

    2005-01-01

    Full text of publication follows: While graphite is widely considered a heat-resistant material, e.g. able to screen metallic shielding from thermal damage, and graphite powder is used as a fire extinguisher agent where water or carbon dioxide should not, it still can react with air and - being carbon - give forth a significant amount of heat. Whether this makes it a hazard in operations such as dismantling nuclear reactors that contain hundreds of tons of graphite, including a small percentage of powder, is a question that has to be answered, considering that dismantling implies the use of such potential fire initiators as thermal cutters and electrical equipment. For this reason EDF commissioned the Centre National de Prevention et Protection (CNPP) to carry out explosibility tests on unirradiated, nuclear grade (i.e. with about 100 ppm of impurities) graphite powder. CNPP tests were so designed as to simulate realistic conditions that might result from a severe mishap during a dismantling operation, such as the crash of heavy equipment on graphite blocks coupled with the bruise of a high power electrical cable. EDF-CNPP tests complement others, done either in Italy most notably on irradiated graphite dust contaminated with various pollutants, or in the UK where the ability of settled graphite dust to propagate an initial gas explosion into an adjacent volume was assessed. EDF-CNPP tests comprise two steps. Step one was intended to produce a qualitative understanding of how nuclear grade graphite behaves while heated in air. In a first series of experiments graphite samples were heated up to 900 C during two and a half hours and their mass loss measured: it was found that while fine powder is wholly oxidised, coarser powder and chunks retained about two thirds of their initial mass. Oxidation kinetics, as assessed by oven temperature shoot-up, begins at 580 C and is quite low, compared with that of iron powder. In a second series of experiments a graphite piece

  3. Oxidizability and explosibility of pure graphite powder

    International Nuclear Information System (INIS)

    Rahmani, L.; Roubineau, D.; Cornet, S.

    2005-01-01

    Full text of publication follows: While graphite is widely considered a heat-resistant material, e.g. able to screen metallic shielding from thermal damage, and graphite powder is used as a fire extinguisher agent where water or carbon dioxide should not, it still can react with air and - being carbon - give forth a significant amount of heat. Whether this makes it a hazard in operations such as dismantling nuclear reactors that contain hundreds of tons of graphite, including a small percentage of powder, is a question that has to be answered, considering that dismantling implies the use of such potential fire initiators as thermal cutters and electrical equipment. For this reason EDF commissioned the Centre National de Prevention et Protection (CNPP) to carry out explosibility tests on unirradiated, nuclear grade (i.e. with about 100 ppm of impurities) graphite powder. CNPP tests were so designed as to simulate realistic conditions that might result from a severe mishap during a dismantling operation, such as the crash of heavy equipment on graphite blocks coupled with the bruise of a high power electrical cable. EDF-CNPP tests complement others, done either in Italy most notably on irradiated graphite dust contaminated with various pollutants, or in the UK where the ability of settled graphite dust to propagate an initial gas explosion into an adjacent volume was assessed. EDF-CNPP tests comprise two steps. Step one was intended to produce a qualitative understanding of how nuclear grade graphite behaves while heated in air. In a first series of experiments graphite samples were heated up to 900 C during two and a half hours and their mass loss measured: it was found that while fine powder is wholly oxidised, coarser powder and chunks retained about two thirds of their initial mass. Oxidation kinetics, as assessed by oven temperature shoot-up, begins at 580 C and is quite low, compared with that of iron powder. In a second series of experiments a graphite piece

  4. Modelling deformation and fracture of Gilsocarbon graphite subject to service environments

    Science.gov (United States)

    Šavija, Branko; Smith, Gillian E.; Heard, Peter J.; Sarakinou, Eleni; Darnbrough, James E.; Hallam, Keith R.; Schlangen, Erik; Flewitt, Peter E. J.

    2018-02-01

    Commercial graphites are used for a wide range of applications. For example, Gilsocarbon graphite is used within the reactor core of advanced gas-cooled reactors (AGRs, UK) as a moderator. In service, the mechanical properties of the graphite are changed as a result of neutron irradiation induced defects and porosity arising from radiolytic oxidation. In this paper, we discuss measurements undertaken of mechanical properties at the micro-length-scale for virgin and irradiated graphite. These data provide the necessary inputs to an experimentally-informed model that predicts the deformation and fracture properties of Gilsocarbon graphite at the centimetre length-scale, which is commensurate with laboratory test specimen data. The model predictions provide an improved understanding of how the mechanical properties and fracture characteristics of this type of graphite change as a result of exposure to the reactor service environment.

  5. 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

  6. Recent developments in graphite

    International Nuclear Information System (INIS)

    Cunningham, J.E.

    1983-01-01

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

  7. 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

  8. Understanding Creep Mechanisms in Graphite with Experiments, Multiscale Simulations, and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Eapen, Jacob [North Carolina State Univ., Raleigh, NC (United States); Murty, Korukonda [North Carolina State Univ., Raleigh, NC (United States); Burchell, Timothy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-06-02

    Disordering mechanisms in graphite have a long history with conflicting viewpoints. Using Raman and x-ray photon spectroscopy, electron microscopy, x-ray diffraction experiments and atomistic modeling and simulations, the current project has developed a fundamental understanding of early-to-late state radiation damage mechanisms in nuclear reactor grade graphite (NBG-18 and PCEA). We show that the topological defects in graphite play an important role under neutron and ion irradiation.

  9. 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.

  10. AGC 2 Irradiation Creep Strain Data Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Windes, William E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rohrbaugh, David T. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    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.

  11. SHMUTZ & PROTON-DIAMANT H + Irradiated/Written-Hyper/Super-conductivity(HC/SC) Precognizance/Early Experiments Connections: Wet-Graphite Room-Tc & Actualized MgB2 High-Tc: Connection to Mechanical Bulk-Moduli/Hardness: Diamond Hydrocarbon-Filaments, Disorder, Nano-Powders:C,Bi,TiB2,TiC

    Science.gov (United States)

    Wunderman, Irwin; Siegel, Edward Carl-Ludwig; Lewis, Thomas; Young, Frederic; Smith, Adolph; Dresschhoff-Zeller, Gieselle

    2013-03-01

    SHMUTZ: ``wet-graphite''Scheike-....[Adv.Mtls.(7/16/12)]hyper/super-SCHMUTZ-conductor(S!!!) = ``wet''(?)-``graphite''(?) = ``graphene''(?) = water(?) = hydrogen(?) =ultra-heavy proton-bands(???) = ...(???) claimed room/high-Tc/high-Jc superconductOR ``p''-``wave''/ BAND(!!!) superconductIVITY and actualized/ instantiated MgB2 high-Tc superconductors and their BCS- superconductivity: Tc Siegel[ICMAO(77);JMMM 7,190(78)] connection to SiegelJ.Nonxline-Sol.40,453(80)] disorder/amorphous-superconductivity in nano-powders mechanical bulk/shear(?)-moduli/hardness: proton-irradiated diamond, powders TiB2, TiC,{Siegel[Semis. & Insuls.5:39,47, 62 (79)])-...``VS''/concommitance with Siegel[Phys.Stat.Sol.(a)11,45(72)]-Dempsey [Phil.Mag. 8,86,285(63)]-Overhauser-(Little!!!)-Seitz-Smith-Zeller-Dreschoff-Antonoff-Young-...proton-``irradiated''/ implanted/ thermalized-in-(optimal: BOTH heat-capacity/heat-sink & insulator/maximal dielectric-constant) diamond: ``VS'' ``hambergite-borate-mineral transformable to Overhauser optimal-high-Tc-LiBD2 in Overhauser-(NW-periodic-table)-Land: CO2/CH4-ETERNAL-sequestration by-product: WATER!!!: physics lessons from

  12. Carbon-14 Graphitization Chemistry

    Science.gov (United States)

    Miller, James; Collon, Philippe; Laverne, Jay

    2014-09-01

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

  13. Graphite Technology Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    W. Windes; T. Burchell; R. Bratton

    2007-09-01

    This technology development plan is designed to provide a clear understanding of the research and development direction necessary for the qualification of nuclear grade graphite for use within the Next Generation Nuclear Plant (NGNP) reactor. The NGNP will be a helium gas cooled Very High Temperature Reactor (VHTR) with a large graphite core. Graphite physically contains the fuel and comprises the majority of the core volume. Considerable effort will be required to ensure that the graphite performance is not compromised during operation. Based upon the perceived requirements the major data needs are outlined and justified from the perspective of reactor design, reatcor performance, or the reactor safety case. The path forward for technology development can then be easily determined for each data need. How the data will be obtained and the inter-relationships between the experimental and modeling activities will define the technology development for graphite R&D. Finally, the variables affecting this R&D program are discussed from a general perspective. Factors that can significantly affect the R&D program such as funding, schedules, available resources, multiple reactor designs, and graphite acquisition are analyzed.

  14. Melting temperature of graphite

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  15. Some equipment for graphite research in swimming pool reactors

    International Nuclear Information System (INIS)

    Seguin, M.; Arragon, Ph.; Dupont, G.; Gentil, J.; Tanis, G.

    1964-01-01

    The irradiation devices described are used for research concerning reactors of the natural uranium type, moderated by graphite and cooled by carbon dioxide. The devices are generally designed for use in swimming pool reactors. The following points have been particularly studied: - maximum use of the irradiation volume, - use of the simplest technological solutions, - standardization of certain constituent parts. This standardization calls for precision machining and careful assembling; these requirements are also true when a relatively low irradiation temperature is required and the nuclear heating is pronounced. Finally, the design of these devices is suitable for the irradiation of other fissile or non-fissile materials. (authors) [fr

  16. Graphitic matrix materials for spherical HTR fuel elements

    International Nuclear Information System (INIS)

    Schulze, R.E.; Schulze, H.A.

    1981-02-01

    The report comprises the graphical documentation of irradiation results on graphitic matrix materials for spherical HTR fuel elements. The plotted results are based on data analyses of the series of exposures in the High Flux Reactor Petten (HFR). The documentation includes information about the changes of - the dimensions - the dynamic modulus of elasticity - the coefficient of thermal expansion of the materials after irradiation with fast neutrons. The irradiation experiments and the data analyses are part of the matrix development and irradiation programme, whose objective, realization and results obtained are summarized. (orig./IHOE) [de

  17. On estimating the fracture probability of nuclear graphite components

    International Nuclear Information System (INIS)

    Srinivasan, Makuteswara

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-05-01

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

  19. Contribution to the study of internal friction in graphites

    International Nuclear Information System (INIS)

    Merlin, J.

    1969-03-01

    A study has been made of the internal friction in different graphites between -180 C and +500 C using a torsion pendulum; the graphites had been previously treated thermo-mechanically, by neutron irradiation and subjected to partial annealings. It has been shown that there occurs: a hysteretic type dissipation of energy, connected with interactions between dislocations and other defects in the matrix; a dissipation having a partially hysteretic character which can be interpreted by a Granato-Luke type formalism and which is connected with the presence of an 'ultra-micro porosity'; a dissipation by a relaxation mechanism after a small dose of irradiation; this is attributed to the reorientation of bi-interstitials; a dissipation having the characteristics of a solid state transformation, this during an annealing after irradiation. It is attributed to the reorganization of interstitial defects. Some information has thus been obtained concerning graphites, in particular: their behaviour at low mechanical stresses, the nature of irradiation defects and their behaviour during annealing, the structural changes occurring during graphitization, the relationship between internal friction and macroscopic mechanical properties. (author) [fr

  20. Cesium diffusion in graphite

    International Nuclear Information System (INIS)

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

    1980-05-01

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

  1. Graphite-based photovoltaic cells

    Science.gov (United States)

    Lagally, Max; Liu, Feng

    2010-12-28

    The present invention uses lithographically patterned graphite stacks as the basic building elements of an efficient and economical photovoltaic cell. The basic design of the graphite-based photovoltaic cells includes a plurality of spatially separated graphite stacks, each comprising a plurality of vertically stacked, semiconducting graphene sheets (carbon nanoribbons) bridging electrically conductive contacts.

  2. Graphite technology development plan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-07-01

    This document presents the plan for the graphite technology development required to support the design of the 350 MW(t) Modular HTGR within the US National Gas-Cooled Reactor Program. Besides descriptions of the required technology development, cost estimates, and schedules, the plan also includes the associated design functions and design requirements.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  4. Uni axial tensile graphite creep capsules with continuous strain registration

    International Nuclear Information System (INIS)

    Hausen, H.; Loelgen, R.; Cundy, M.

    1977-01-01

    Two irradiation devices are described for the in pile measurement of tensile irradiation creep of graphite. In each machine a single sample is maintained under a controlled load by a pneumatic bellows system. Irradiation creep is measured relative to unstressed reference shells which surround the stressed sample. This differential strain is detected by linear displacement transducers, and recorded automatically by the out of pile installation. Irradiation temperatures are in the 800 to 1100 0 C range, and the stresses up to 60% of the U.T.S. One machine has been specifically designed for a flux change experiment, other irradiation parameters remaining fixed. Temperature control is achieved through varying gas mixtures in control gas gaps. The paper details the design principles of the machines and gives an account of the cold and hot commissioning tests, with particular reference to the accuracy of the in pile measuring system. Finally, the early irradiation experience is evaluated

  5. Modification of PMMA/graphite nanocomposites through ion beam technique

    Science.gov (United States)

    Singhal, Prachi; Rattan, Sunita; Avasthi, Devesh Kumar; Tripathi, Ambuj

    2013-08-01

    Swift heavy ion (SHI) irradiation is a special technique for inducing physical and chemical modifications in bulk materials. In the present work, the SHI hs been used to prepare nanocomposites with homogeneously dispersed nanoparticles. The nanographite was synthesized from graphite using the intercalation-exfoliation method. PMMA Poly(methyl methacrylate)/graphite nanocomposites have been synthesized by in situ polymerization. The prepared PMMA/graphite nanocomposite films were irradiated with SHI irradiation (Ni ion beam, 80 MeV and C ion beam, 50 MeV) at a fluence of 1×1010 to 3×1012 ions/cm2. The nanocomposite films were characterized by scanning electron microscope (SEM) and were evaluated for their electrical and sensor properties. After irradiation, significant changes in surface morphology of nanocomposites were observed as evident from the SEM images, which show the presence of well-distributed nanographite platelets. The irradiated nanocomposites exhibit better electrical and sensor properties for the detection of nitroaromatics with marked improvement in sensitivity as compared with unirradiated nanocomposites.

  6. A German research project about applicable graphite cutting techniques

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  7. Harwell Graphite Calorimeter

    International Nuclear Information System (INIS)

    Linacre, J.K.

    1970-01-01

    The calorimeter is of the steady state temperature difference type. It contains a graphite sample supported axially in a graphite outer jacket, the assembly being contained in a thin stainless steel outer can. The temperature of the jacket and the temperature difference between sample and jacket are measured by chromel-alumel thermocouples. The instrument is calibrated by means of an electric heater of low mass positioned on the axis of the sample. The resistance of the heater is known and both current through the heater and the potential across it may be measured. The instrument is filled with nitrogen at a pressure of one half atmosphere at room temperature. The calorimeter has been designed for prolonged operation at temperatures up to 600°C, and dose rates up to 1 Wg -1 , and instruments have been in use for periods in excess of one year

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

    Energy Technology Data Exchange (ETDEWEB)

    Le Guillou, M. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon, F-69622 Villeurbanne cedex (France); Moncoffre, N., E-mail: n.moncoffre@ipnl.in2p3.fr [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon, F-69622 Villeurbanne cedex (France); Toulhoat, N. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon, F-69622 Villeurbanne cedex (France); CEA/DEN – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Pipon, Y. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon, F-69622 Villeurbanne cedex (France); Institut Universitaire Technologique, Université Claude Bernard Lyon 1, Université de Lyon, F-69622 Villeurbanne cedex (France); Ammar, M.R. [CNRS, CEMHTI UPR3079, Université Orléans, CS90055, F-45071 Orléans cedex 2 (France); Rouzaud, J.N.; Deldicque, D. [Laboratoire de Géologie de l’Ecole Normale Supérieure, Paris, UMR CNRS ENS 8538, F-75231 Paris cedex 5 (France)

    2016-03-15

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

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

    Science.gov (United States)

    Le Guillou, M.; Moncoffre, N.; Toulhoat, N.; Pipon, Y.; Ammar, M. R.; Rouzaud, J. N.; Deldicque, D.

    2016-03-01

    This paper is a contribution to the study of the behavior of activation products produced in irradiated nuclear graphite, graphite being the moderator of the first French generation of CO2 cooled nuclear fission reactors. This paper is focused on the thermal release of Tritium, a major contributor to the initial activity, taking into account the role of the free surfaces (open pores and graphite surface). Two kinds of graphite were compared. On one hand, Highly Oriented Pyrolitic Graphite (HOPG), a model well graphitized graphite, and on the other hand, SLA2, a porous less graphitized nuclear graphite. Deuterium ion implantation at three different energies 70, 200 and 390 keV allows simulating the presence of Tritium at three different depths, corresponding respectively to projected ranges Rp of 0.75, 1.7 and 3.2 μm. The D isotopic tracing is performed thanks to the D(3He,p)4He nuclear reaction. The graphite structure is studied by Raman microspectrometry. Thermal annealing is performed in the temperature range 200-1200 °C up to 300 h annealing time. As observed in a previous study, the results show that the D release occurs according to three kinetic regimes: a rapid permeation through open pores, a transient regime corresponding to detrapping and diffusion of D located at low energy sites correlated to the edges of crystallites and finally a saturation regime attributed to detrapping of interstitial D located at high energy sites inside the crystallites. Below 600 °C, D release is negligible whatever the implantation depth and the graphite type. The present paper clearly puts forward that above 600 °C, the D release decreases at deeper implantation depths and strongly depends on the graphite structure. In HOPG where high energy sites are more abundant, the D release is less dependent on the surface proximity compared to SLA2. In SLA2, in which the low energy sites prevail, the D release curves are clearly shifted towards lower temperatures when D is located

  10. Micro-orientation control of silicon polymer thin films on graphite surfaces modified by heteroatom doping

    Energy Technology Data Exchange (ETDEWEB)

    Shimoyama, Iwao, E-mail: shimoyama.iwao@jaea.go.jp [Material Science Research Center, Atomic Energy Agency, Tokai-mura 2-4, Naka-gun, Ibaraki 319-1195 (Japan); Baba, Yuji [Fukushima Administrative Department, Atomic Energy Agency, Tokai-mura 2-4, Naka-gun, Ibaraki 319-1195 (Japan); Hirao, Norie [Material Science Research Center, Atomic Energy Agency, Tokai-mura 2-4, Naka-gun, Ibaraki 319-1195 (Japan)

    2017-05-31

    Highlights: • Micro-orientation control method for organic polysilane thin films is proposed. • This method utilizes surface modification of graphite using heteroatom doping. • Lying, standing, and random orientations can be freely controlled by this method. • Micro-pattering of a polysilane film with controlled orientations is achieved. - Abstract: Near-edge X-ray absorption fine structure (NEXAFS) spectroscopy is applied to study orientation structures of polydimethylsilane (PDMS) films deposited on heteroatom-doped graphite substrates prepared by ion beam doping. The Si K-edge NEXAFS spectra of PDMS show opposite trends of polarization dependence for non irradiated and N{sub 2}{sup +}-irradiated substrates, and show no polarization dependence for an Ar{sup +}-irradiated substrate. Based on a theoretical interpretation of the NEXAFS spectra via first-principles calculations, we clarify that PDMS films have lying, standing, and random orientations on the non irradiated, N{sub 2}{sup +}-irradiated, and Ar{sup +}-irradiated substrates, respectively. Furthermore, photoemission electron microscopy indicates that the orientation of a PDMS film can be controlled with microstructures on the order of μm by separating irradiated and non irradiated areas on the graphite surface. These results suggest that surface modification of graphite using ion beam doping is useful for micro-orientation control of organic thin films.

  11. Structural disorder of graphite and implications for graphite thermometry

    Directory of Open Access Journals (Sweden)

    M. Kirilova

    2018-02-01

    Full Text Available Graphitization, or the progressive maturation of carbonaceous material, is considered an irreversible process. Thus, the degree of graphite crystallinity, or its structural order, has been calibrated as an indicator of the peak metamorphic temperatures experienced by the host rocks. However, discrepancies between temperatures indicated by graphite crystallinity versus other thermometers have been documented in deformed rocks. To examine the possibility of mechanical modifications of graphite structure and the potential impacts on graphite thermometry, we performed laboratory deformation experiments. We sheared highly crystalline graphite powder at normal stresses of 5 and 25  megapascal (MPa and aseismic velocities of 1, 10 and 100 µm s−1. The degree of structural order both in the starting and resulting materials was analyzed by Raman microspectroscopy. Our results demonstrate structural disorder of graphite, manifested as changes in the Raman spectra. Microstructural observations show that brittle processes caused the documented mechanical modifications of the aggregate graphite crystallinity. We conclude that the calibrated graphite thermometer is ambiguous in active tectonic settings.

  12. GRAFEC: A New Spanish Program to Investigate Waste Management Options for Radioactive Graphite - 12399

    Energy Technology Data Exchange (ETDEWEB)

    Marquez, Eva; Pina, Gabriel; Rodriguez, Marina [CIEMAT, Av. Complutense, 22, 28040-MADRID (Spain); Fachinger, Johannes; Grosse, Karl-Heinz [Furnaces Nuclear Application Grenoble SAS (FNAG), 4, avenue Charles de Gaulle, 38800 Le Pont de Claix (France); Leganes Nieto, Jose Luis; Quiros Gracian, Maria [ENRESA, C/ Emilio Vargas,7 - 28043 - MADRID (Spain); Seemann, Richard [ALD Vacuum Technologies GmbH, Wilhelm-Rohn-Strasse 35, 63450 Hanau (Germany)

    2012-07-01

    Spain has to manage about 3700 tons of irradiated graphite from the reactor Vandellos I as radioactive waste. 2700 tons are the stack of the reactor and are still in the reactor core waiting for retrieval. The rest of the quantities, 1000 tons, are the graphite sleeves which have been already retrieved from the reactor. During operation the graphite sleeves were stored in a silo and during the dismantling stage a retrieval process was carried out separating the wires from the graphite, which were crushed and introduced into 220 cubic containers of 6 m{sup 3} each and placed in interim storage. The graphite is an intermediate level radioactive waste but it contains long lived radionuclides like {sup 14}C which disqualifies disposal at the low level waste repository of El Cabril. Therefore, a new project has been started in order to investigate two new options for the management of this waste type. The first one is based on a selective decontamination of {sup 14}C by thermal methods. This method is based on results obtained at the Research Centre Juelich (FZJ) in the Frame of the EC programs 'Raphael' and 'Carbowaste'. The process developed at FZJ is based on a preferential oxidation of {sup 14}C in comparison to the bulk {sup 12}C. Explanations for this effect are the inhomogeneous distribution and a weaker bounding of {sup 14}C which is not incorporated in the graphite lattice. However these investigations have only been performed with graphite from the high temperature reactor Arbeitsgemeinschaft Versuchsreaktor Juelich AVR which has been operated in a non-oxidising condition or research reactor graphite operated at room temperature. The reactor Vandellos I has been operated with CO{sub 2} as coolant and significant amounts of graphite have been already oxidised. The aim of the project is to validate whether a {sup 14}C decontamination can also been achieved with graphite from Vandellos I. A second possibility under investigation is the

  13. Bromine intercalated graphite for lightweight composite conductors

    KAUST Repository

    Amassian, Aram

    2017-07-20

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

  14. Chemical stabilization of graphite surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Bistrika, Alexander A.; Lerner, Michael M.

    2018-04-03

    Embodiments of a device, or a component of a device, including a stabilized graphite surface, methods of stabilizing graphite surfaces, and uses for the devices or components are disclosed. The device or component includes a surface comprising graphite, and a plurality of haloaryl ions and/or haloalkyl ions bound to at least a portion of the graphite. The ions may be perhaloaryl ions and/or perhaloalkyl ions. In certain embodiments, the ions are perfluorobenzenesulfonate anions. Embodiments of the device or component including stabilized graphite surfaces may maintain a steady-state oxidation or reduction surface current density after being exposed to continuous oxidation conditions for a period of at least 1-100 hours. The device or component is prepared by exposing a graphite-containing surface to an acidic aqueous solution of the ions under oxidizing conditions. The device or component can be exposed in situ to the solution.

  15. Impedance of electrochemically modified graphite.

    Science.gov (United States)

    Magdić, Katja; Kvastek, Krešimir; Horvat-Radošević, Višnja

    2014-01-01

    Electrochemical impedance spectroscopy, EIS, has been applied for characterization of electrochemically modified graphite electrodes in the sulphuric acid solution. Graphite modifications were performed by potential cyclization between potentials of graphite oxide formation/reduction, different number of cycles, and prolonged reduction steps after cyclization. Impedance spectra measured at two potential points within double-layer region of graphite have been successfully modeled using the concept of porous electrodes involving two different electrolyte diffusion paths, indicating existence of two classes of pores. The evaluated impedance parameter values show continuous changes with stages of graphite modification, indicating continuous structural changes of pores by number of potential cycles applied. Differences of impedance parameter values at two potential values indicate the potential induced changes of solution properties within the pores of modified graphite.

  16. 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.

  17. A model for pulsed laser melting of graphite

    Science.gov (United States)

    Steinbeck, J.; Braunstein, G.; Dresselhaus, M. S.; Venkatesan, T.; Jacobson, D. C.

    1985-12-01

    A model for laser melting of carbon at high temperatures to form liquid carbon has been developed. This model is solved numerically using experimental data from laser irradiation studies in graphite consistent with a melting temperature for graphite of 4300 K. The parameters for high-temperature graphite are based on the extension of previously measured thermal properties into the high-temperature regime. A simple classical free electron gas model is used to calculate the properties of liquid carbon. There is very good agreement between the model calculation and experimental results for laser pulse fluences below 2.0 J/cm2. Modifications to the model for larger laser pulse fluences are discussed.

  18. Heat exchanger using graphite foam

    Science.gov (United States)

    Campagna, Michael Joseph; Callas, James John

    2012-09-25

    A heat exchanger is disclosed. The heat exchanger may have an inlet configured to receive a first fluid and an outlet configured to discharge the first fluid. The heat exchanger may further have at least one passageway configured to conduct the first fluid from the inlet to the outlet. The at least one passageway may be composed of a graphite foam and a layer of graphite material on the exterior of the graphite foam. The layer of graphite material may form at least a partial barrier between the first fluid and a second fluid external to the at least one passageway.

  19. Blunt indentation of core graphite

    International Nuclear Information System (INIS)

    Hartley, M.; McEnaney, B.

    1996-01-01

    Blunt indentation experiments were carried out on unoxidized and thermally oxidised IM1-24 graphite as a model to simulate local point stresses acting on graphite moderator bricks. Blunt indentation of unoxidized graphite initiates cracks close to the region of maximum tensile stress at the edge of the indentation. Cracks propagate and converge to form a cone of material. Failure is catastrophic, typically forming three pieces of graphite and ejecting the cone referred to above. The failure mode under indentation loading for highly oxidised graphite (weigh loss > 40%) is different from that for the unoxidized graphite. There is no longer a distinct crack path, the indentation is much deeper than in the case of the unoxidized graphite, and there is a region of crushed debris beneath the indentation, producing a crater-like structure. The reduction in the compressive fracture stress, σ cf , under indentation loading with increasing fractional weight loss on oxidation, x, can be fitted to σ cf /σ 0 = exp-[5.2x] where σ 0 is the compressive fracture stress of the unoxidized graphite. This indicates that the effect of thermal oxidation on indentation fracture stress is more severe than the effects of radiolytic oxidation on conventional strengths of nuclear graphites. (author). 8 refs, 12 figs

  20. Mathematical modelling of the the processes of radiating formation of defects at interaction of carbon with graphite

    Directory of Open Access Journals (Sweden)

    A. Kupchishin

    2012-12-01

    Full Text Available Processes of radiation formation of defects in the carbon irradiated by graphite are considered in work. The regularities arising at selection of approximation expressions, a finding of result area at calculation of cascadely - probabilistic functions depending on number of interactions and depth of penetration of particles are revealed. The regularities formed at calculations of concentration of radiating defects in graphite, irradiated by carbon are received. Results of calculations are presented in the form of tables and schedules.

  1. Irradiation techniques for carbon materials (Osiris reactor)

    International Nuclear Information System (INIS)

    Genthon, J.P.; Micaud, G.; Mottet, P.

    1976-01-01

    Neutron irradiation devices are described for carbons, graphites and fuel compacts. The use of standard COLIBRI furnaces of different diameters in the OSIRIS core allows long-life irradiations to be carried out, with or without stress, in situ measurements, or fission gas samplings. Examples of the obtained results are given [fr

  2. New irradiation devices at the FRN reactor

    International Nuclear Information System (INIS)

    Stark, W.

    1980-01-01

    In order to fulfill the experimental demands three additional devices were constructed and installed. The first is a vertical irradiation tube in air surrounded by a lead cylinder (in the irradiation position). The second device is a rabbit system ending within the graphite moderator of the thermal column. The third device is so called rotating disk assembly, built to replace the rotary specimen rack

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

    International Nuclear Information System (INIS)

    Rahmani, L.

    2016-01-01

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

  4. Graphitic packing removal tool

    Science.gov (United States)

    Meyers, Kurt Edward; Kolsun, George J.

    1997-01-01

    Graphitic packing removal tools for removal of the seal rings in one piece. he packing removal tool has a cylindrical base ring the same size as the packing ring with a surface finish, perforations, knurling or threads for adhesion to the seal ring. Elongated leg shanks are mounted axially along the circumferential center. A slit or slits permit insertion around shafts. A removal tool follower stabilizes the upper portion of the legs to allow a spanner wrench to be used for insertion and removal.

  5. Heavy-ion irradiation induced diamond formation in carbonaceous materials

    International Nuclear Information System (INIS)

    Daulton, T. L.

    1999-01-01

    The basic mechanisms of metastable phase formation produced under highly non-equilibrium thermodynamic conditions within high-energy particle tracks are investigated. In particular, the possible formation of diamond by heavy-ion irradiation of graphite at ambient temperature is examined. This work was motivated, in part, by earlier studies which discovered nanometer-grain polycrystalline diamond aggregates of submicron-size in uranium-rich carbonaceous mineral assemblages of Precambrian age. It was proposed that the radioactive decay of uranium formed diamond in the fission particle tracks produced in the carbonaceous minerals. To test the hypothesis that nanodiamonds can form by ion irradiation, fine-grain polycrystalline graphite sheets were irradiated with 400 MeV Kr ions. The ion irradiated graphite (and unirradiated graphite control) were then subjected to acid dissolution treatments to remove the graphite and isolate any diamonds that were produced. The acid residues were then characterized by analytical and high-resolution transmission electron microscopy. The acid residues of the ion-irradiated graphite were found to contain ppm concentrations of nanodiamonds, suggesting that ion irradiation of bulk graphite at ambient temperature can produce diamond

  6. Nonlinear seismic analysis of a graphite reactor core

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  7. 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.

  8. 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)

  9. 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)

  10. Investigation of carbon near the graphite-diamond-liquid triple point

    International Nuclear Information System (INIS)

    Prawer, S.; Jamieson, D.N.

    1992-01-01

    Pulsed laser irradiation is used to heat deeply buried damage layers in diamond. Over a small range of laser powers, damage annealing, formation of buried graphitic layers, and melting of diamond followed by its conversion upon cooling into graphite are observed. The diagnostics employed are Channeling Contrast Microscopy, optical absorption, surface profilometry, and scanning and optical microscopies. The results are explained in terms of the behaviour of carbon under high internal pressures close to the diamond-graphite-liquid carbon triple point in the phase diagram. 17 refs., 3 figs

  11. Surface analysis of graphite fiber reinforced polyimide composites

    Science.gov (United States)

    Messick, D. L.; Progar, D. J.; Wightman, J. P.

    1983-01-01

    Several techniques have been used to establish the effect of different surface pretreatments on graphite-polyimide composites. Composites were prepared from Celion 6000 graphite fibers and the polyimide LARC-160. Pretreatments included mechanical abrasion, chemical etching and light irradiation. Scanning electron microscopy (SEM) and X-ray photoelectron spectroscopy (XPS) were used in the analysis. Contact angle of five different liquids of varying surface tensions were measured on the composites. SEM results showed polymer-rich peaks and polymer-poor valleys conforming to the pattern of the release cloth used durng fabrication. Mechanically treated and light irradiated samples showed varying degrees of polymer peak removal, with some degradation down to the graphite fibers. Minimal changes in surface topography were observed on concentrations of surface fluorine even after pretreatment. The light irradiation pretreatment was most effective at reducing surface fluorine concentrations whereas chemical pretreatment was the least effective. Critical surface tensions correlated directly with the surface fluorine to carbon ratios as calculated from XPS.

  12. AGC-1 Post Irradiation Examination Status

    Energy Technology Data Exchange (ETDEWEB)

    David Swank

    2011-09-01

    The Next Generation Nuclear Plant (NGNP) Graphite R&D program is currently measuring irradiated material property changes in several grades of nuclear graphite for predicting their behavior and operating performance within the core of new Very High Temperature Reactor (VHTR) designs. The Advanced Graphite Creep (AGC) experiment consisting of six irradiation capsules will generate this irradiated graphite performance data for NGNP reactor operating conditions. All six AGC capsules in the experiment will be irradiated in the Advanced Test Reactor (ATR), disassembled in the Hot Fuel Examination Facility (HFEF), and examined at the INL Research Center (IRC) or Oak Ridge National Laboratory (ORNL). This is the first in a series of status reports on the progress of the AGC experiment. As the first capsule, AGC1 was irradiated from September 2009 to January 2011 to a maximum dose level of 6-7 dpa. The capsule was removed from ATR and transferred to the HFEF in April 2011 where the capsule was disassembled and test specimens extracted from the capsules. The first irradiated samples from AGC1 were shipped to the IRC in July 2011and initial post irradiation examination (PIE) activities were begun on the first 37 samples received. PIE activities continue for the remainder of the AGC1 specimen as they are received at the IRC.

  13. Acoustic emission from polycrystalline graphites

    International Nuclear Information System (INIS)

    Ioka, I.; Yoda, S.; Oku, T.; Miyamoto, Y.

    1987-01-01

    Acoustic emission was monitored from polycrystalline graphites with different microstructure (pore size and pore volume) subjected to compressive loading. The graphites used in this study comprised five brands, that is, PGX, ISEM-1, IG-11, IG-15, and ISO-88. A root mean square (RMS) voltage and event counts of acoustic emission for graphites were measured during compressive loading. The acoustic emission was measured using a computed-based data acquisition and analysis system. The graphites were first deformed up to 80 % of the average fracture stress, then unloaded and reloaded again until the fracture occured. During the first loading, the change in RMS voltage for acoustic emission was detected from the initial stage. During the unloading, the RMS voltage became zero level as soon as the applied stress was released and then gradually rose to a peak and declined. The behavior indicated that the reversed plastic deformation occured in graphites. During the second loading, the RMS voltage gently increased until the applied stress exceeded the maximum stress of the first loading; there is no Kaiser effect in the graphites. A bicrystal model could give a reasonable explanation of this results. The empirical equation between the ratio of σ AE to σ f and σ f was obtained. It is considered that the detection of microfracture by the acoustic emission technique is effective in macrofracture prediction of polycrystalline graphites. (author)

  14. Hypervelocity impacts into graphite

    Science.gov (United States)

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

    2011-03-01

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

  15. Radiolytic graphite oxidation revisited

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    Science.gov (United States)

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

    2017-11-01

    Different matrix and graphite morphologies were generated by a special heat treatment in three chemically different series of flake graphite cast iron samples. As cast, furnace cooled and air cooled samples were investigated. The length of graphite particles and the pearlite volume of samples were determined by metallographic examination and these parameters were compared with the nondestructively measured magnetic parameters. Magnetic measurements were performed by the method of Magnetic Adaptive Testing, which is based on systematic measurement and evaluation of minor magnetic hysteresis loops. It was shown that linear correlation existed between the magnetic quantities and the graphite length, and also between the magnetic quantities and the relative pearlite content in the investigated cast iron. A numerical expression was also determined between magnetic descriptors and relative pearlite content, which does not depend on the detailed experimental conditions.

  17. Eutectic solidification mode of spheroidal graphite cast iron and graphitization

    Directory of Open Access Journals (Sweden)

    Hideo Nakae

    2007-02-01

    Full Text Available The shrinkage and chilling tendency of spheroidal graphite (abbreviated SG cast iron is much greater than that of the flake graphite cast iron in spite of its higher amount of C and Si contents. Why? The main reason should be the difference in their graphitization during the eutectic solidification. In this paper, we discuss the difference in the solidification mechanism of both cast irons for solving these problems using unidirectional solidification and the cooling curves of the spheroidal graphite cast iron. The eutectic solidification rate of the SG cast iron is controlled by the diffusion of carbon through the austenite shell, and the final thickness is 1.4 times the radius of the SG, therefore, the reduction of the SG size, namely, the increase in the number, is the main solution of these problems.

  18. 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

  19. Study of graphitic microstructure formation in diamond bulk by pulsed Bessel beam laser writing

    Science.gov (United States)

    Kumar, S.; Sotillo, B.; Chiappini, A.; Ramponi, R.; Di Trapani, P.; Eaton, S. M.; Jedrkiewicz, O.

    2017-11-01

    The advantages of using Bessel beams for the generation of graphitic structures in diamond bulk are presented. We show that by irradiating the sample with a pulsed Bessel beam whose non-diffracting zone is of the same order of the sample thickness, it is possible to produce without any sample translation straight graphitic through-microstructures, whose size depends on the input pulse energy. The microstructure growth is investigated as a function of the number of irradiating pulses, and the femtosecond and picosecond regimes are contrasted.

  20. CALANDRIA TYPE SODIUM GRAPHITE REACTOR

    Science.gov (United States)

    Peterson, R.M.; Mahlmeister, J.E.; Vaughn, N.E.; Sanders, W.J.; Williams, A.C.

    1964-02-11

    A sodium graphite power reactor in which the unclad graphite moderator and fuel elements are contained within a core tank is described. The core tank is submersed in sodium within the reactor vessel. Extending longitudinally through the core thnk are process tubes with fuel elements positioned therein. A bellows sealing means allows axial expansion and construction of the tubes. Within the core tank, a leakage plenum is located below the graphite, and above the graphite is a gas space. A vent line regulates the gas pressure in the space, and another line removes sodium from the plenum. The sodium coolant flows from the lower reactor vessel through the annular space between the fuel elements and process tubes and out into the reactor vessel space above the core tank. From there, the heated coolant is drawn off through an outlet line and sent to the heat exchange. (AEC)

  1. 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

  2. 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)

  3. Graphite oral tattoo: case report

    OpenAIRE

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

    2015-01-01

    Pigmented oral lesions compose a large number of pathological entities, including exogenous pigmentat oral tattoos, such as amalgam and graphite tattoos. We report a rare case of a graphite tattoo on the palate of a 62-year-old patient with a history of pencil injury, compare it with amalgam tattoos, and determine the prevalence of oral tattoos in our Oral Pathology Service. We also compare the clinical and histological findings of grafite and amalgam tattoos. Oral tattoos affect women more f...

  4. Modelling fracture of aged graphite bricks under radiation and temperature

    Directory of Open Access Journals (Sweden)

    Atheer Hashim

    2017-05-01

    Full Text Available The graphite bricks of the UK carbon dioxide gas cooled nuclear reactors are subjected to neutron irradiation and radiolytic oxidation during operation which will affect thermal and mechanical material properties and may lead to structural failure. In this paper, an empirical equation is obtained and used to represent the reduction in the thermal conductivity as a result of temperature and neutron dose. A 2D finite element thermal analysis was carried out using Abaqus to obtain temperature distribution across the graphite brick. Although thermal conductivity could be reduced by up to 75% under certain conditions of dose and temperature, analysis has shown that it has no significant effect on the temperature distribution. It was found that the temperature distribution within the graphite brick is non-radial, different from the steady state temperature distribution used in the previous studies [1,2]. To investigate the significance of this non-radial temperature distribution on the failure of graphite bricks, a subsequent mechanical analysis was also carried out with the nodal temperature information obtained from the thermal analysis. To predict the formation of cracks within the brick and the subsequent propagation, a linear traction–separation cohesive model in conjunction with the extended finite element method (XFEM is used. Compared to the analysis with steady state radial temperature distribution, the crack initiation time for the model with non-radial temperature distribution is delayed by almost one year in service, and the maximum crack length is also shorter by around 20%.

  5. Oxidation Protective SiC Coating on Graphite for VHTR Core Support Structure

    International Nuclear Information System (INIS)

    Park, Jae-Won; Kim, Eung-Seon; Kim, Jae-Un; Windes, William E.

    2014-01-01

    The potential for reducing oxidation of the supporting graphite components during normal operation and accident conditions in the VHTR design has been studied. SiC coating on graphite has been studied taking into consideration of possible dimensional change of graphite by the neutron-irradiation. Functionally gradient (FG) SiC coating on the graphite has been performed to moderate the SiC/Graphite interface: E-beam evaporative coating from varied compositions of graphite/SiC mixture in the source crucibles was carried out with an ion beam mixing. The cylindrical graphite samples were uniformly coated by rotating and revolving the samples. Auger depth profile reveals that the ion beam mixed interface is broadened and a cross sectional EDS Si elemental mapping shows a smoothly graded Si profile. The grown film exhibited a stacked columnar structure owing to a frequent sample position change during the coating process, as observed by FE-SEM. As a result of 18 thermal cycling test of 500-1000℃, no film delamination was found on the coated layer, but film cracks were formed, suggesting a strong bonding. When samples were heated at 600°C in static air for 2 h, ~45 wt% of the graphite was burnt off, whereas for the SiC coated graphite only 5 wt %. When heated at 1000 °C in air, vigorous oxidation of graphite took place through a few paths (maybe the mars and/or the crack lines) in the film only leaving the coating layer. As the crack lines were covered with SiC by repeating the ion beam mixed coating process, the oxidation resistance was improved. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  8. Graphite Formation in Cast Iron

    Science.gov (United States)

    Stefanescu, D. M.

    1985-01-01

    In the first phase of the project it was proven that by changing the ratio between the thermal gradient and the growth rate for commercial cast iron samples solidifying in a Bridgman type furnace, it is possible to produce all types of graphite structures, from flake to spheroidal, and all types of matrices, from ferritic to white at a certain given level of cerium. KC-135 flight experiments have shown that in a low-gravity environment, no flotation occurs even in spheroidal graphite cast irons with carbon equivalent as high as 5%, while extensive graphite flotation occurred in both flake and spheroidal graphite cast irons, in high carbon samples solidified in a high gravity environment. This opens the way for production of iron-carbon composite materials, with high carbon content (e.g., 10%) in a low gravity environment. By using KC-135 flights, the influence of some basic elements on the solidification of cast iron will be studied. The mechanism of flake to spheroidal graphite transition will be studied, by using quenching experiments at both low and one gravity for different G/R ratios.

  9. 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

  10. Nondestructive evaluation method on mechanical property change of graphite components in the HTGR by ultrasonic wave propagation with grain/pore microstructure

    International Nuclear Information System (INIS)

    Shibata, Taiju; Ishihara, Masahiro

    2003-01-01

    Oxidation damage is one of the crucial factors to degrade mechanical properties of graphite components in the HTGRs. The oxidation increases the porosity of graphite and, hence, results in degradation. In order to evaluate the oxidation damage at neutron irradiated conditions, a new analytical method by ultrasonic wave propagation characteristics was developed. Irradiation effects, a dimensional change and a pinning of dislocations in crystals, on the propagation characteristics in graphite are taken into consideration in the method. It was shown that an equivalent velocity of the wave in graphite is increased by the irradiation, and that a signal height of a propagated waveform is increased by the irradiation, and it decreases with increasing porosity caused by the oxidation. The Young's modulus for an ideal graphite polycrystals without pore was evaluated by considering the wave velocity in them in order to evaluate the change of the apparent modulus at simultaneous irradiated and oxidized conditions as an application of the developed method. It was also shown that the oxidation-induced change of the modulus is appropriately evaluated by the method, suggesting that it is possible to evaluate the change for the irradiated conditions. It can be said from this study that the developed method is promising to evaluate the oxidation damage on graphite components in the HTGRs by nondestructive way. (author)

  11. Energy response of graphite-mixed magnesium borate TLDs to low energy x-rays

    DEFF Research Database (Denmark)

    Pelliccioni, M.; Prokic, M.; Esposito, A.

    1991-01-01

    Graphite-mixed sintered magnesium borate TL dosemeters are attractive for beta/gamma dosimetry because they combine a low energy dependence to beta-rays with near tissue or air equivalence to photon irradiations and a high sensitivity. In this paper results from the experimental measurements...

  12. Mechanical properties of graphites and carbon materials

    International Nuclear Information System (INIS)

    Jouquet, Gilbert.

    1977-01-01

    The mechanical behavior of graphites and artificial carbons is related to the structure of these materials. The influence of structural modifications in a graphite monocrystal on the deformation and fracture properties is studied [fr

  13. RECOVERY OF VALUABLE MATERIAL FROM GRAPHITE BODIES

    Science.gov (United States)

    Fromm, L.W. Jr.

    1959-09-01

    An electrolytic process for recovering uranium from a graphite fuel element is described. The uraniumcontaining graphite body is disposed as the anode of a cell containing a nitric acid electrolyte and a 5 amp/cm/sup 2/ current passed to induce a progressive disintegration of the graphite body. The dissolved uranium is quickly and easily separated from the resulting graphite particles by simple mechanical means, such as centrifugation, filtration, and decontamination.

  14. Residual stress measurements in polycrystalline graphite with micro-Raman spectroscopy

    International Nuclear Information System (INIS)

    Krishna, Ram; Jones, Abbie N.; Edge, Ruth; Marsden, Barry J.

    2015-01-01

    Micro-Raman microscopy technique is applied to evaluate unevenly distributed residual stresses in the various constituents of polygranular reactor grades graphite. The wavenumber based Raman shift (cm −1 ) corresponds to the local residual stress and measurements of stress dependent first order Raman spectra in graphite have enabled localized residual stress values to be determined. The bulk polygranular graphite of reactor grades – Gilsocarbon, NBG-18 and PGA – are examined to illustrate the residual stress variations in their constituents. Binder phase and filler particles have shown to be under compressive and tensile stresses, respectively. Among the studied graphite grades, the binder phase in Gilsocarbon has the highest residual stress and NBG-18 has the lowest value. Filler particles in Gilsocarbon have the highest residual stress and PGA showed the lowest, this is most likely due to the morphology of the coke particles used in the manufacturing and applied processing techniques for fabrications. Stresses have also been evaluated along the peripheral of pores and at the tips of the cracks. Cracks in filler and binder phases have shown mixed behaviour, compressive as well as tensile, whereas pores in binder and filler particles have shown compressive behaviour. The stresses in these graphitic constituents are of the order of MPa. Non-destructive analyses presented in this study make the current state-of-the-art technique a powerful method for the study of stress variations near the graphite surface and are expected to increase its use further in property determination analysis of low to highly fluence irradiated graphite samples from the material test reactors. - Highlights: • Micro-Raman spectroscopy can measure significantly small residual stresses. • Gilsocarbon, NBG-18 and PGA graphite were evaluated for residual stresses. • Residual stresses in the constituents of graphite were evaluated. • Binder and filler particles are often found under

  15. 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

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

    International Nuclear Information System (INIS)

    Sumita, Junya; Shimazaki, Yosuke; Shibata, Taiju

    2014-01-01

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

  17. Superconductivity in graphite intercalation compounds

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-15

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

  18. Graphite nanoreinforcements in polymer nanocomposites

    Science.gov (United States)

    Fukushima, Hiroyuki

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

  19. Study of high dose nitrogen implantation into graphite

    International Nuclear Information System (INIS)

    Romanovskij, E.A.; Bespalova, O.V.; Borisov, A.M.; Goryaga, N.G.; Zatekin, V.V.; Kulikauskas, V.S.; Sukharev, V.G.

    1997-01-01

    Rutherford backscattering spectroscopy was used for the study of high dose (35 keV)N + ions implantation into graphites and glassy carbon. Quantitative date on depth profiles and its dependences on irradiation fluence and ion beam flux were obtained for all elements. The stationary cupola-shaped depth profile with maximum nitrogen concentration 22-27% (at.) is reached at sufficiently large fluence. The obtained results are discussed in the frame of high dose implantation models and compared with results of another methods of carbon nitride synthesis

  20. Laser ultrasonic assessment of the effects of porosity and microcracking on the elastic moduli of nuclear graphites

    International Nuclear Information System (INIS)

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

    2016-01-01

    Laser ultrasonic methods have been used to measure the elastic moduli of various nuclear graphites. Measurements were made to assess wavespeeds for longitudinal and shear waves in different propagation directions and these were used along with density measurements to compute the longitudinal and shear moduli as well as Young's modulus. All moduli decreased with increasing graphite porosity and these variations could be interpreted using models describing the effect of porosity on material stiffness. Extrapolations for these models to zero porosity were used to infer the moduli for theoretically dense graphite; the results were far below predicted values reported in the literature for fully dense, polycrystalline, isotropic graphite. Differences can be attributed to microcracking in the graphite microstructure. Using models for the effects of microcracking on modulus, estimates for microcrack populations indicate that the number of cracks per unit volume must be much greater than the number of pores per unit volume. Experimental results reported in the literature for irradiated graphites as well as for the stress dependence of graphite modulus are consistent with the influence of microcracking on elastic behavior and could be interpreted using concepts developed here. Results in this work for graphite structure-property relationships should allow for more sophisticated characterization of nuclear graphites using ultrasonic methods. - Highlights: • Moduli of nuclear graphites measured using laser ultrasonic methods. • Estimates made for the moduli of fully dense, polycrystalline, isotropic graphite. • Models for modulus used to assess effects of microcracking on stiffness. • Microcrack densities estimated using models for effect of microcracks on modulus.

  1. Seismic Study of TMSR Graphite Core Structure

    International Nuclear Information System (INIS)

    Tsang, Derek; Huang Chao Chao

    2014-01-01

    Graphite plays an important role in the thorium based molten salt reactor (TMSR) nuclear energy system. The graphite core acts as reflector, moderator and structural material in the TMSR core. The graphite core assembly has hundreds of graphite bricks interconnected with graphite keys and dowels. In other words, the graphite core is a kind of discrete stack structure with highly nonlinear dynamic behaviour, and it will show totally different dynamics responses comparing with welded structure or bolted structure when subjected to the seismic loading. Hence it is important to investigate the dynamics characteristics of the TMSR graphite core assembly and to meet the seismic design requirement. The most popular way to investigate the nonlinearity of graphite core is to do finite element analyses. Due to the large number of nonlinear behaviour caused by contacts, collisions and impacts between the graphite bricks and keys, the computational costs on seismic analysis of the whole core would be very high. Many methods have been developed in the past 20 years to conquer this difficulty. In this work substructure method and finite element method have been used to study the dynamic behaviour of a stack of graphite bricks under seismic loading. The numerical results of these two methods will be compared. The results show that the super element method is an efficient method for graphite core seismic analyses. (author)

  2. The effects of space radiation on a chemically modified graphite-epoxy composite material

    Science.gov (United States)

    Reed, S. M.; Herakovich, C. T.; Sykes, G. F.

    1986-01-01

    The effects of the space environment on the engineering properties and chemistry of a chemically modified T300/934 graphite-epoxy composite system are characterized. The material was subjected to 1.0 x 10 to the 10th power rads of 1.0 MeV electron irradiation under vacuum to simulate 30 years in geosynchronous earth orbit. Monotonic tension tests were performed at room temperature (75 F/24 C) and elevated temperature (250 F/121 C) on 4-ply unidirectional laminates. From these tests, inplane engineering and strength properties (E sub 1, E sub 2, Nu sub 12, G sub 12, X sub T, Y sub T) were determined. Cyclic tests were also performed to characterize energy dissipation changes due to irradiation and elevated temperature. Large diameter graphite fibers were tested to determine the effects of radiation on their stiffness and strength. No significant changes were observed. Dynamic-mechanical analysis demonstrated that the glass transition temperature was reduced by 50 F(28 C) after irradiation. Thermomechanical analysis showed the occurrence of volatile products generated upon heating of the irradiated material. The chemical modification of the epoxy did not aid in producing a material which was more radiation resistant than the standard T300/934 graphite-epoxy system. Irradiation was found to cause crosslinking and chain scission in the polymer. The latter produced low molecular weight products which plasticize the material at elevated temperatures and cause apparent material stiffening at low stresses at room temperature.

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  4. Evaluation of co-cokes from bituminous coal with vacuum resid or decant oil, and evaluation of anthracites, as precursors to graphite

    Science.gov (United States)

    Nyathi, Mhlwazi S.

    2011-12-01

    Graphite is utilized as a neutron moderator and structural component in some nuclear reactor designs. During the reactor operaction the structure of graphite is damaged by collision with fast neutrons. Graphite's resistance to this damage determines its lifetime in the reactor. On neutron irradiation, isotropic or near-isotropic graphite experiences less structural damage than anisotropic graphite. The degree of anisotropy in a graphite artifact is dependent on the structure of its precursor coke. Currently, there exist concerns over a short supply of traditional precursor coke, primarily due to a steadily increasing price of petroleum. The main goal of this study was to study the anisotropic and isotropic properties of graphitized co-cokes and anthracites as a way of investigating the possibility of synthesizing isotropic or near-isotropic graphite from co-cokes and anthracites. Demonstrating the ability to form isotropic or near-isotropic graphite would mean that co-cokes and anthracites have a potential use as filler material in the synthesis of nuclear graphite. The approach used to control the co-coke structure was to vary the reaction conditions. Co-cokes were produced by coking 4:1 blends of vacuum resid/coal and decant oil/coal at temperatures of 465 and 500 °C for reaction times of 12 and 18 hours under autogenous pressure. Co-cokes obtained were calcined at 1420 °C and graphitized at 3000 °C for 24 hours. Optical microscopy, X-ray diffraction, temperature-programmed oxidation and Raman spectroscopy were used to characterize the products. It was found that higher reaction temperature (500 °C) or shorter reaction time (12 hours) leads to an increase in co-coke structural disorder and an increase in the amount of mosaic carbon at the expense of textural components that are necessary for the formation of anisotropic structure, namely, domains and flow domains. Characterization of graphitized co-cokes showed that the quality, as expressed by the degree of

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

    Science.gov (United States)

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

    2017-09-11

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

  6. 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.

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

    Czech Academy of Sciences Publication Activity Database

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

    2017-01-01

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

  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. Self-propagating solar light reduction of graphite oxide in water

    Energy Technology Data Exchange (ETDEWEB)

    Todorova, N.; Giannakopoulou, T.; Boukos, N.; Vermisoglou, E. [Institute of Nanoscience and Nanotechnology, NCSR “Demokritos”, 153 41 Attikis (Greece); Lekakou, C. [Division of Mechanical, Medical, and Aerospace Engineering, Faculty of Engineering and Physical Sciences, University of Surrey, Guildford (United Kingdom); Trapalis, C., E-mail: c.trapalis@inn.demokritos.gr [Institute of Nanoscience and Nanotechnology, NCSR “Demokritos”, 153 41 Attikis (Greece)

    2017-01-01

    Highlights: • Graphite oxide was partially reduced by solar light irradiation in water media. • No addition of catalysts nor reductive agent were used for the reduction. • Specific capacitance increased stepwise with increase of irradiation time. • Self-propagating reduction of graphene oxide by solar light is suggested. - Abstract: Graphite Oxide (GtO) is commonly used as an intermediate material for preparation of graphene in the form of reduced graphene oxide (rGO). Being a semiconductor with tunable band gap rGO is often coupled with various photocatalysts to enhance their visible light activity. The behavior of such rGO-based composites could be affected after prolonged exposure to solar light. In the present work, the alteration of the GtO properties under solar light irradiation is investigated. Water dispersions of GtO manufactured by oxidation of natural graphite via Hummers method were irradiated into solar light simulator for different periods of time without addition of catalysts or reductive agent. The FT-IR analysis of the treated dispersions revealed gradual reduction of the GtO with the increase of the irradiation time. The XRD, FT-IR and XPS analyses of the obtained solid materials confirmed the transition of GtO to rGO under solar light irradiation. The reduction of the GtO was also manifested by the CV measurements that revealed stepwise increase of the specific capacitance connected with the restoration of the sp{sup 2} domains. Photothermal self-propagating reduction of graphene oxide in aqueous media under solar light irradiation is suggested as a possible mechanism. The self-photoreduction of GtO utilizing solar light provides a green, sustainable route towards preparation of reduced graphene oxide. However, the instability of the GtO and partially reduced GO under irradiation should be considered when choosing the field of its application.

  10. A 3-D inelastic analysis of HTR graphite structures and a comparison with A 2-D approach

    International Nuclear Information System (INIS)

    Willaschek, J.

    1979-01-01

    In High Temperature Reactor Cores (HTR) a large number of elements are constructed of nuclear graphite. The dimensions of the graphite components are limited by stresses and strains resulting from thermal loads, irradiation induced dimensional changes and stress-dependent irradiation creep. Therefore it is necessary to examine the feasibility of design concepts with regard to the structural integrity of the material. This paper presents an analysis of a radial reflector concept for use in a 3000 MWth HTR for process heat production. This concept of a pebble bed reactor (OTTO cycle) requires reflector dimensions and shapes which have previously not been used and which may exceed acceptable stress limits. Graphite reflector elements in a HTR are subject to a high fluence of fast neutrons. The fluence varies spatially within an element. Irradiation-induced strains occur which in turn vary non-linearly with the fluence. At low fluences the graphite shrinks. With increasing fluence shrinkage is saturated and after a 'turn-around' point the graphite begins to swell. The net effect of fluence gradient and irradiation-induced strain is a 'necking' of the element which moves radially outwards with time. In this paper a three-dimensional inelastic analysis of a graphite block with the above deformation history is described. The influence of irradiation on dimensional stability and other material properties was taken into account. Numerical results were obtained with the finite-element computer code ADINA, modified at INTERATOM for the task in hand. The radial reflector block was modelled using 21-node three-dimensional continuum elements of elastic-creep material. The element stiffness matrices were calculated using the standard 2x2x2 Gauss integration; material nonlinearities with quadratic displacement functions and linearised initial strains were employed. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Zhamu, Aruna; Jang, Bor Z.

    2014-06-17

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

  12. HRB-22 irradiation phase test data report

    International Nuclear Information System (INIS)

    Montgomery, F.C.; Acharya, R.T.; Baldwin, C.A.; Rittenhouse, P.L.; Thoms, K.R.; Wallace, R.L.

    1995-03-01

    Irradiation capsule HRB-22 was a test capsule containing advanced Japanese fuel for the High Temperature Test Reactor (HTTR). Its function was to obtain fuel performance data at HTTR operating temperatures in an accelerated irradiation environment. The irradiation was performed in the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL). The capsule was irradiated for 88.8 effective full power days in position RB-3B of the removable beryllium (RB) facility. The maximum fuel compact temperature was maintained at or below the allowable limit of 1300 degrees C for a majority of the irradiation. This report presents the data collected during the irradiation test. Included are test thermocouple and gas flow data, the calculated maximum and volume average temperatures based on the measured graphite temperatures, measured gaseous fission product activity in the purge gas, and associated release rate-to-birth rate (R/B) results. Also included are quality assurance data obtained during the test

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

    Energy Technology Data Exchange (ETDEWEB)

    Burchell, Timothy D [ORNL; Bratton, Rob [Idaho National Laboratory (INL); Marsden, Barry [University of Manchester, UK; Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission; Penfield, Scott [Technology Insights; Mitchell, Mark [PBMR (Pty) Ltd.; Windes, Will [Idaho National Laboratory (INL)

    2008-03-01

    ) concepts, such as the NGNP, it is fully expected that the behavior of these graphites will conform to the recognized trends for near isotropic nuclear graphite. Thus, much of the data needed is confirmatory in nature. Theories that can explain graphite behavior have been postulated and, in many cases, shown to represent experimental data well. However, these theories need to be tested against data for the new graphites and extended to higher neutron doses and temperatures pertinent to the new Gen IV reactor concepts. It is anticipated that current and planned future graphite irradiation experiments will provide the data needed to validate many of the currently accepted models, as well as providing the needed data for design confirmation.

  14. Pyrolytic graphite gauge for measuring heat flux

    Science.gov (United States)

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

    2002-01-01

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

  15. Uranium Oxide Aerosol Transport in Porous Graphite

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-01-23

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

  16. ISX-A graphite limiter experiment

    International Nuclear Information System (INIS)

    Langley, R.A.; Colchin, R.J.; Isler, R.C.; Murakami, M.; Simpkins, J.E.; Cecchi, J.L.; Corso, V.L.; Dylla, H.F.; Ellis, R.A. Jr.; Nishi, M.

    1979-01-01

    Graphite limiters were installed and tested in the ISX-A tokamak as part of the ISX-A surface physics program and the TFTR materials research program. The puropse of the experiment was to compare plasma performance using graphite limiters as opposed to the standard ISX-A stainless steel limiters. Heaters were installed in the graphite limiters so that the effects of operation at elevated temperatures could be evaluated

  17. Graphitic Carbon Nitride Materials for Energy Applications

    OpenAIRE

    Belen Jorge, A.; Dedigama, I.; Mansor, N.; Jervis, R.; Corà, F.; McMillan, P. F.; Brett, D.

    2015-01-01

    Polymeric layered carbon nitrides were investigated for use as catalyst support materials for proton exchange membrane fuel cells (PEMFCs) and water electrolyzers (PEMWEs). Three different carbon nitride materials were prepared: a heptazine-based graphitic carbon nitride material (gCNM), poly (triazine) imide carbon nitride intercalated with LiCl component (PTI-Li+Cl-) and boron-doped graphitic carbon nitride (B-gCNM). Following accelerated corrosion testing, all graphitic carbon nitride mate...

  18. Dynamics of graphite flake on a liquid

    Science.gov (United States)

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

    2006-11-01

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

  19. Graphite-to-metal bonding techniques

    International Nuclear Information System (INIS)

    Lindquist, L.O.; Mah, R.

    1977-11-01

    The results of various bonding methods to join graphite to different metals are reported. Graphite/metal bonds were tested for thermal flux limits and thermal flux cycling lifetimes. The most successful bond transferred a heat flux of 6.50 MW/m 2 in more than 500 thermal cycles. This bond was between pyrolytic graphite and copper with Ti-Cu-Sil as the bonding agent

  20. Effect of heating temperature on the structure and fine texture of carbon-graphite pan fibres

    International Nuclear Information System (INIS)

    Varenkov, A.N.

    1976-01-01

    The structure and strength properties of carbon-graphite fibres obtained at 1000, 2000, 2750 and 3000 0 C from polyacrylonitrile (PAN) were investigated. The dependence of the interplanar distance, dimensions of the quasicrystallites and their orientations on the fibre temperature obtained were determined by X-ray structural analysis. A rise in formation temperature of the fibre leads to a decreased interplanar distance, (from 6.031 to 6.798 A), to growth of the quasicrystallites (from 14 to 99 A) and to a decrease in their degree of disorientation with respect to the fibre axis. Re-annealing influences the structure and properties of the fibres in the case where the temperature of the re-anneal exceeds the temperature at which the carbon-graphite fibre is obtained. Bombardment with neutrons ( E > 0.5 MeV) lowers the strength of the carbon-graphite fibres as a function of the irradiation dose. (author)

  1. AC induction field heating of graphite foam

    Science.gov (United States)

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

    2017-08-22

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

  2. Characterization of commercial expandable graphite fire retardants

    Energy Technology Data Exchange (ETDEWEB)

    Focke, Walter Wilhelm, E-mail: walter.focke@up.ac.za; Badenhorst, Heinrich; Mhike, Washington; Kruger, Hermanus Joachim; Lombaard, Dewan

    2014-05-01

    Highlights: • Expandable graphite is less well-ordered than its graphite bisulfate progenitor. • It includes graphite oxide as a randomly interstratified phase. • CO{sub 2}, CO and SO{sub 2} are released during thermal-driven exfoliation. - Abstract: Thermal analysis and other techniques were employed to characterize two expandable graphite samples. The expansion onset temperatures of the expandable graphite's were ca. 220 °C and 300 °C respectively. The key finding is that the commercial products are not just pure graphite intercalation compounds with sulfuric acid species intercalated as guest ions and molecules in between intact graphene layers. A more realistic model is proposed where graphite oxide-like layers are also randomly interstratified in the graphite flakes. These graphite oxide-like layers comprise highly oxidized graphene sheets which contain many different oxygen-containing functional groups. This model explains the high oxygen to sulfur atomic ratios found in both elemental analysis of the neat materials and in the gas generated during the main exfoliation event.

  3. 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

  4. Study of corrosion resistance graphite in oxygen

    International Nuclear Information System (INIS)

    Zelenskij, V.F.; Odejchuk, N.P.; Petel'guzov, I.A.; Ryzhov, V.P.; Yakovlev, V.K.

    2011-01-01

    The paper presents the results of the corrosion resistance of MPG, ARV and GSP graphite grades in oxygen at temperatures of 400, 600 and 800 o C. The oxidation kinetics of graphites is defined. It is shown, that interaction process of graphites with oxygen is characterized by a decrease of sample weights. The description of installation for carrying out of tests and a technique of carrying out of tests and researches is resulted. It is shown that the best corrosion resistance in the investigated temperature range has GSP graphite with density of 1.8-1.9 g/cm 3 of NSC KIPT production.

  5. Low temperature vapor phase digestion of graphite

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, Robert A.

    2017-04-18

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

  6. International Comparison on Thermal-Diffusivity Measurements for Iron and Isotropic Graphite Using the Laser Flash Method in CCT-WG9

    Science.gov (United States)

    Akoshima, M.; Hay, B.; Zhang, J.; Chapman, L.; Baba, T.

    2013-05-01

    The first international pilot study of thermal-diffusivity measurements using the laser flash (LF) method was organized by the working group 9 (WG9) of the Consultative Committee for Thermometry (CCT) of the Bureau International des Poids et Mesures (BIPM). Four National Metrology Institutes (NMIs) participated in this comparison. Thermal-diffusivity measurements on the Armco iron and the isotropic graphite IG-110 were carried out from room temperature to about 1200 K. The sample sets consist of five disk-shaped specimens of 10 mm in diameter and (1.0, 1.4, 2.0, 2.8, and 4.0) mm in thickness, each cut from the same block of material. These sample sets were specifically prepared for the comparison and sent to the participants. In the pilot comparison, the thermal diffusivity of each sample was estimated using the LF method with a specific extrapolating procedure. This procedure has the advantage of determining the inherent thermal diffusivity of the material. The extrapolated value in a plot of measured apparent thermal-diffusivity values versus the amplitude of the output signal corresponding to the temperature rise during each measurement is defined as the inherent thermal diffusivity. The overall results showed good agreement between independent laboratories, measurement equipment, and specimen thicknesses. The thermal diffusivities of the materials were determined using our measured results. A quantitative evaluation of the variability of the data obtained by the participants has been done, by evaluating the deviations from the reference value, the Z-value, and the En-number. Some data showed a large deviation from the reference value. It was concluded that these are caused by an insufficient time response of the measurement equipment and some difficulties with changing the pulsed heating energy. The effect of the thermal expansion on the thermal diffusivity was checked. It was found that the thermal-expansion effect was very small and negligible in this case.

  7. Estimation of radioactivity of graphite blocks in Tokai Power Station using statistical method

    International Nuclear Information System (INIS)

    Nakano, Masaaki; Mikami, Hisashi; Ichige, Hideaki; Tsukada, Shinichi

    2011-01-01

    Tokai Power Station (graphite moderated, gas-cooled reactor, GCR) stopped its commercial operation in March 1998 and is decommissioning now. Since graphite blocks in Tokai reactor core are major low level wastes (LLWs), the realistic and reasonable method to estimate radioactivity of graphite blocks is required for final disposal and its licensing procedure. In general, LLWs, which were installed in or around a reactor core, have large radioactivity, theoretical calculations can be applied to the estimation of the radioactivity. This paper describes the concept of the method using statistical approach to determine the radioactivity of the graphite blocks in the reactor core. This method directly considers the variations of input calculation conditions, for example, compositions of impurity elements, irradiation neutron flux and irradiation period. In this paper, the variations of the compositions of impurity elements were statistically considered with the mean value and the standard deviation that were determined with element analyses. Many activation calculations were performed with the compositions that were determined with pseudorandom numbers, the mean value and the standard deviation. The calculated radioactivities distribute also statistically and a mean value and a standard deviation of radioactivity can be determined. The distribution of calculated radioactivities shows consistency to radiochemical analyses of graphite blocks from the reactor core and this shows that the method is applicable to the estimation of the graphite block radioactivity. Furthermore, this method can be considered to reduce over-excess estimation margin and can obtain reasonable radioactivity rather than using maximum or conservative values of all input conditions. This method is now being developed and approved as one of basic procedure for determining the radioactivity of wastes by Standards Committee of the Atomic Energy Society of Japan. (author)

  8. Examination of Surface Deposits on Oldbury Reactor Core Graphite to Determine the Concentration and Distribution of 14C.

    Directory of Open Access Journals (Sweden)

    Liam Payne

    Full Text Available Pile Grade A graphite was used as a moderator and reflector material in the first generation of UK Magnox nuclear power reactors. As all of these reactors are now shut down there is a need to examine the concentration and distribution of long lived radioisotopes, such as 14C, to aid in understanding their behaviour in a geological disposal facility. A selection of irradiated graphite samples from Oldbury reactor one were examined where it was observed that Raman spectroscopy can distinguish between underlying graphite and a surface deposit found on exposed channel wall surfaces. The concentration of 14C in this deposit was examined by sequentially oxidising the graphite samples in air at low temperatures (450°C and 600°C to remove the deposit and then the underlying graphite. The gases produced were captured in a series of bubbler solutions that were analysed using liquid scintillation counting. It was observed that the surface deposit was relatively enriched with 14C, with samples originating lower in the reactor exhibiting a higher concentration of 14C. Oxidation at 600°C showed that the remaining graphite material consisted of two fractions of 14C, a surface associated fraction and a graphite lattice associated fraction. The results presented correlate well with previous studies on irradiated graphite that suggest there are up to three fractions of 14C; a readily releasable fraction (corresponding to that removed by oxidation at 450°C in this study, a slowly releasable fraction (removed early at 600°C in this study, and an unreleasable fraction (removed later at 600°C in this study.

  9. A review of the behaviour of graphite under the conditions appropriate for protection of the first wall of a fusion reactor

    International Nuclear Information System (INIS)

    Birch, M.; Brocklehurst, J.E.

    1987-12-01

    The material used as a first wall protection in fusion reactor systems will be exposed to 14 MeV neutrons from the fusion reaction and suffer surface bombardment by other energetic particles in the plasma. Graphite is a potential candidate for the first wall material. Calculations are performed of the damaging power of 14 MeV neutrons so that existing graphite irradiation data can be utilised. Such data at high irradiation temperatures are reviewed for a wide range of graphite types, characterised by specific examples, and the application of the data to design calculations is discussed. The erosion/corrosion effect of the plasma at the graphite surface is also considered. Limitations in the state of knowledge are identified, and particular areas of further work are recommended. (author)

  10. Food irradiation

    International Nuclear Information System (INIS)

    Soothill, R.

    1987-01-01

    The issue of food irradiation has become important in Australia and overseas. This article discusses the results of the Australian Consumers' Association's (ACA) Inquiry into food irradiation, commissioned by the Federal Government. Issues discussed include: what is food irradiation; why irradiate food; how much food is consumer rights; and national regulations

  11. Food irradiation

    International Nuclear Information System (INIS)

    Munasiri, M.A.

    1990-01-01

    Gives details of sources used for food irradiation, brief description of the process, safety of food irradiation process, practical applications and the amount of doses used for spices, condiments, mangoes etc., limitations of food irradiation, international status of clearance of irradiated foods, versatility of the process

  12. In situ atomic-scale observation of irradiation induced carbon nanocrystalline formation from dense carbon clusters

    Science.gov (United States)

    Wang, Chengbing; Li, Zhengtong; Ling, San; Lei, Tao; Su, Jinbu

    2018-03-01

    We present a direct observation of the transformation of dense amorphous carbon clusters into diamond nanocrystalline under electron beam irradiation by in situ transmission electron microscopy, where the surrounding carbon matrix did not significantly change. Our findings provide clear and convincing evidence for the diamond nanocrystalline evolving from energetic amorphous carbon sites. Furthermore, graphitization of amorphous carbons usually demands a high temperature combined with high pressure. Hence, graphitization of amorphous carbons at relatively low temperatures is highly desired. Here we offer a useful method for catalyst-free graphitization of amorphous carbons by employing moderate electron beam irradiation, without external heating being applied.

  13. Laser surface graphitization to control friction of diamond-like carbon coatings

    Science.gov (United States)

    Komlenok, Maxim S.; Kononenko, Vitaly V.; Zavedeev, Evgeny V.; Frolov, Vadim D.; Arutyunyan, Natalia R.; Chouprik, Anastasia A.; Baturin, Andrey S.; Scheibe, Hans-Joachim; Shupegin, Mikhail L.; Pimenov, Sergei M.

    2015-11-01

    To study the role of laser surface graphitization in the friction behavior of laser-patterned diamond-like carbon (DLC) films, we apply the scanning probe microscopy (SPM) in the lateral force mode (LFM) which allows to obtain simultaneously the lateral force and topography images and to determine local friction levels in laser-irradiated and original surface areas. Based on this approach in the paper, we report on (1) laser surface microstructuring of hydrogenated a-C:H and hydrogen-free ta-C films in the regime of surface graphitization using UV laser pulses of 20-ns duration and (2) correlation between the structure and friction properties of the laser-patterned DLC surface on micro/nanoscale using SPM/LFM technique. The SPM/LFM data obtained for the surface relief gratings of graphitized microstructures have evidenced lower friction forces in the laser-graphitized regions. For the hydrogenated DLC films, the reversible frictional behavior of the laser-graphitized micropatterns is found to take place during LFM imaging at different temperatures (20 and 120 °C) in ambient air. It is revealed that the lateral force distribution in the laser-graphitized areas is shifted to higher friction levels (relative to that of the unirradiated surface) at temperature 120 °C and returned back to the lower friction during the sample cooling to 20 °C, thus confirming an influence of adsorbed water layers on the nanofriction properties of laser-graphitized micropatterns on the film surface.

  14. Characteristics of dissimilar laser-brazed joints of isotropic graphite to WC-Co alloy

    Energy Technology Data Exchange (ETDEWEB)

    Nagatsuka, Kimiaki, E-mail: nagatuka@jwri.osaka-u.ac.jp [Graduate School of Engineering, Osaka University, Joining and Welding Research Institute, 11-1, Mihogaoka, Ibaraki, Osaka 567-0047 (Japan); Sechi, Yoshihisa, E-mail: sechi@kagoshima-it.go.jp [Kagoshima Prefectural Institute of Industrial Technology, 1445-1 Oda, Hayato-cho, Kirishima, Kagoshima 899-5105 (Japan); Miyamoto, Yoshinari, E-mail: y_miyamoto@toyotanso.co.jp [Toyo Tanso Co., Ltd., 5-7-12 Takeshima, Nishiyodgawa-ku, Osaka 555-0011 (Japan); Nakata, Kazuhiro, E-mail: nakata@jwri.osaka-u.ac.jp [Joining and Welding Research Institute, Osaka University, 11-1, Mihogaoka, Ibaraki, Osaka 567-0047 (Japan)

    2012-04-25

    Highlights: Black-Right-Pointing-Pointer Ti was required in the filler metal for brazing graphite to WC-Co alloy. Black-Right-Pointing-Pointer The shear strength of the joint increased with Ti content up to 1.7 mass%. Black-Right-Pointing-Pointer Ti concentrated at the interface of graphite/filler metal. Black-Right-Pointing-Pointer TiC was formed at the interface of graphite/filler metal. - Abstract: The effect of Ti serving as an activator in a eutectic Ag-Cu alloy filler metal in dissimilar laser-brazed joints of isotropic graphite and a WC-Co alloy on the joint strength and the interface structure of the joint is investigated in this study. To evaluate the joint characteristics, the Ti content in the filler metal was increased from 0 to 2.8 mass%. The laser brazing was carried out by irradiating a laser beam selectively on the WC-Co alloy plate in Ar atmosphere. The threshold content of Ti required to join isotropic graphite to WC-Co alloy was 0.4 mass%. The shear strength at the brazed joint increased rapidly with increasing Ti content up to 1.7 mass%, and a higher Ti content was found to be likely to saturate the shear strength to a constant value of about 14 MPa. The isotropic graphite blocks also fractured at this content. The concentration of Ti observed at the interface between isotropic graphite and the filler metal indicates the formation of an intermetallic layer of TiC.

  15. Positron annihilation study of graphite, glassy carbon and C60/C70 fullerene

    International Nuclear Information System (INIS)

    Hasegawa, Masayuki; Kajino, Masahiro; Yamaguchi, Sadae; Iwata, Tadao; Kuramoto, Eiichi; Takenaka, Minoru.

    1992-01-01

    ACAR (Angular Correlation of Annihilation Radiation) and positron lifetime measurements have been made on, HOPG (Highly Oriented Pyrolytic Graphite), isotropic fine-grained graphite, glassy carbons and C 60 /C 70 powder. HOPG showed marked bimodality along the c-axis and anisotropy in ACAR momentum distribution, which stem from characteristic annihilation between 'interlayer' positrons and π-electrons in graphite. ACAR curves of the isotropic graphite and glassy carbons are even narrower than that of HOPG perpendicular to the c-axis. Positron lifetime of 420 and 390 - 480 psec, much longer than that of 221 psec in HOPG, were observed for the isotropic graphite and glassy carbons respectively, which are due to positron trapping in structural voids in them. Positron lifetime and ACAR width (FWHM) can be well correlated to void sizes (1.7 to 5.0 nm) of glassy carbons which have been determined by small angle neutron (SAN) scattering measurements. ACAR curves and positron lifetime of C 60 /C 70 powder agree well with those of glassy carbons. This shows that positron wave functions extend, as in the voids of glassy carbons, much wider than open spaces of the octahedral interstices of the face-centered cubic (FCC) structure of C 60 crystal and strongly suggests positron trapping in the 'soccer ball' vacancy. Possible positron states in the carbon materials are discussed with a simple model of void volume-trapping. Preliminary results on neutron irradiation damage in HOPG are also presented. (author)

  16. 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.)

  17. HTR Fuel Waste Management: TRISO separation and acid-graphite intercalation compounds preparation

    International Nuclear Information System (INIS)

    Guittonneau, Fabrice; Abdelouas, Abdesselam; Grambow, Bernd

    2010-01-01

    Considering the need to reduce waste production and greenhouse emissions and still keeping high energy efficiency, various 4th generation nuclear energy systems have been proposed. As far as graphite-moderated reactors are concerned (future high temperature fast or thermal reactors), one of the key issues is the large volumes of irradiated graphite encountered. With the objective to reduce volume of waste in the HTR concept, it is very important to be able to separate the fuel from low level activity graphite representing a large volume. The separated TRISO particles can then be reprocessed for waste separation or disposed off in geological repository. In addition, preparation of acid-GICs from the separated graphite may constitute a way to recycle this waste. We used HTR-type compact fuel with ZrO 2 TRISO particles to test two separation methods: low (H 2 SO 4 + H 2 O 2 ) and high (H 2 SO 4 + HNO 3 ) temperature acid treatments. In both cases the TRISO separation was complete but some TRISO layers oxidized at high temperature. At low temperature, the desegregation of graphite grains is facilitated by intercalation of sulfuric acid between the graphene layers. The acid-GIC obtained consists of pure phases of high quality suggesting their potential industrial recycling.

  18. Distribution of fission products in graphite sleeves and blocks of the eleventh and twelfth OGL-1 fuel assemblies

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Fukuda, Kousaku; Kikuchi, Teruo; Tsuruta, Harumichi.

    1994-06-01

    The 11th and 12th fuel assemblies were irradiated in an in-pile gas loop, OGL-1, installed in the Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Research Institute (JAERI). Distribution of fission products in the graphite sleeves and blocks of the assemblies was measured by gamma-ray spectrometry. The 11th fuel assembly was aimed at testing the irradiation performance of mass product fuels in trial manufacturing of the first charge fuel for the High Temperature Engineering Test Reactor (HTTR) in relatively short irradiation, and the 12th assembly in long-term irradiation. The 12th assembly attained a burnup approximately as high as that of the HTTR driver fuel design. In the graphite sleeve of the 11th assembly, high concentration peaks of fission products were found in the axial distribution. Exposure of failed fuel particles was not detected on the surface of fuel compacts, while fissures of graphite matrix at overcoat boundaries were observed on the surface. These results led to a presumption that fission products, which were released from failed particles located inside of the fuel compact, was easily transported through the fissures of the matrix to the inner surface of the sleeve. In the graphite sleeve of the 12th assembly, 110m Ag was detected together with other fission products of 137 Cs, 134 Cs etc. Silver-110m showed characteristic distribution: this nuclides was less concentrated at the region with highly concentrated 60 Co which is presumed to have been transported from melted sheath material of thermocouples to the graphite sleeve. It was inferred from the distribution that the transport behavior of 110m Ag had been influenced by co-sorption or by pore structure change in the graphite material of the sleeve, which had been induced by metallic elements including cobalt. (author)

  19. Food irradiation

    International Nuclear Information System (INIS)

    Lindqvist, H.

    1996-01-01

    This paper is a review of food irradiation and lists plants for food irradiation in the world. Possible applications for irradiation are discussed, and changes induced in food from radiation, nutritional as well as organoleptic, are reviewed. Possible toxicological risks with irradiated food and risks from alternative methods for treatment are also brought up. Ways to analyze weather food has been irradiated or not are presented. 8 refs

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

    Science.gov (United States)

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

    2018-05-01

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

  1. Flexible PVC flame retarded with expandable graphite

    CSIR Research Space (South Africa)

    Focke, WW

    2014-02-01

    Full Text Available The utility of expandable graphite as a flame retardant for PVC, plasticized with 60 phr of a phosphate ester, was investigated. Cone calorimeter results, at a radiant flux of 35 kW m 2, revealed that adding only 5 wt.% expandable graphite lowered...

  2. Methane generated from graphite--tritium interaction

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  3. 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

  4. Modeling Fission Product Sorption in Graphite Structures

    Energy Technology Data Exchange (ETDEWEB)

    Szlufarska, Izabela [University of Wisconsin, Madison, WI (United States); Morgan, Dane [University of Wisconsin, Madison, WI (United States); Allen, Todd [University of Wisconsin, Madison, WI (United States)

    2013-04-08

    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

  5. GROPPER, an output program for results from BERSAFE. Phase 3 for graphite

    International Nuclear Information System (INIS)

    Harper, P.G.

    1980-05-01

    BERSAFE, the CEGB's Finite Element Stress Analysis System has been extended to enable analyses to be performed on structures made from irradiated graphite. As with all non-linear analyses there is a considerable amount of output data produced which needs to be selectively processed. GROPPER reads the tape file produced by such an analysis and outputs selected results to the line printer as well as preparing tapes for plotting of results. (author)

  6. Vandellos 1 Graphite Solubilisation and Activity Determination of 36Cl, 99Tc and 129I

    International Nuclear Information System (INIS)

    Rodríguez Alcalá, Marina; Piña Lucas, Gabriel; López Gutiérrez, Jose María; Leganés Nieto, Jose Luis

    2016-01-01

    During the characterization processes of irradiate graphite performed in the dismantling of the UNGG NPP of Vandellós 1, and as a consequence of the sample treatment carried out for measuring the main radionuclides of interest included in the El Cabril inventory, it was not possible to measure the activity of 99 Tc, 129 I and 36 Cl. Further R&D projects have been developed to solve this issue and finally a proper sample treatment has been accomplished. (author)

  7. Water-cooled pyrolytic graphite targets at LAMPF: design and operation

    International Nuclear Information System (INIS)

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

    1981-01-01

    Design considerations and actual operating experience are reported for water-cooled pyrolytic graphite targets at the Clinton P. Anderson Meson Physics Facility (LAMPF). Emphasis is placed on the use of finite element computer calculations to determine target temperatures and stresses, which can then be evaluated to judge the usefulness of a particular design. Consideration is also given to the swelling of the target following irradiation, and to the measures taken to prolong target lifetime

  8. The fracture of graphite; La rupture des graphites

    Energy Technology Data Exchange (ETDEWEB)

    Rouby, D. [Institut National des Sciences Appliquees (INSA), Groupe d' Etudes de Metallurgie Physique et de Physique des Materiaux, UMR CNRS 5510, 69 - Villeurbanne (France); Monchaux, St. [Institut National des Sciences Appliquees (INSA), Dept. Science et Genie des Materiaux, 69 - Villeurbanne (France); Tahon, B. [Laboratoire SGL Carbon SAS, 74 - Passy (France)

    2006-03-15

    By mechanical loading, the behaviour of poly-granular graphites for industrial uses is globally brittle: when a pre-existing flaw becomes critical a crack initiates and then propagates more or less catastrophically. This scheme implies several features which are described in the present paper. First, as the crack will be initiated at a critical flaw, the ultimate stress appears as largely dispersed and the strength is not an intrinsic material's parameter. Secondly, the processing route introduces in the material some microstructure anisotropy, largely influencing the strength dispersion. Finally, the crack propagation is controlled by a bridging mechanism of the lips which depends on the microstructure. This effect can be described by the so-called crack growth resistance curve: the R-curve. (authors)

  9. Damaging process of graphite - new model and its impact on degradation of materials performance

    International Nuclear Information System (INIS)

    Tanabe, T.; Muto, S.

    1999-01-01

    The most widely accepted model for development of defect structure in neutron irradiated graphite has been such that following the first production of a pair of an interstitial and vacancy, di-interstitials and vacancies are formed and their subsequent growth would result in the production of an interstitial plane or loop in-between the basal planes and vacancy clusters, respectively, which could cause the loss of thermal conductivity and dimensional change. Recently we have claimed that the formation of vacancy clusters and growth of the interstitial planes are not necessarily a unique interpretation of the damaging process. Instead, the damaging process is described by orientational disordering within the basal planes, i.e. fragmentation into small crystallites and rotation of their crystalline axes, change of stacking order and elongation of the interplanar spacing. The orientational disordering within the basal planes proceeds coordinately over a few layers with their layered correlation maintained. This process accompanies changes in bonding nature producing 5 member- and 7 member-atomic rings as appeared in fullerenes. This is so to speak ''self-restoring or reconstruction'' to maintain resonance bonds as strict as possible without the formation of dangling bonds. This paper reviews irradiation effects in graphite such as increase of hydrogen retention, loss of thermal conductivity and dimensional change on the bases of our new model, taking account of the changes of the bonding nature in irradiated graphite. (orig.)

  10. Deep levels of nitrogen vacancies complexes in graphite-like boron nitride

    CERN Document Server

    Grinyaev, S N; Lopatin, V V

    2002-01-01

    Paper presents results of theoretical studies using methods of model pseudopotential and of extended elementary cell of deep levels of nitrogen vacancies, small clusters from di- and trivacancies of nitrogen covering nearest defects in one layer of graphite-like boron nitride. On the basis of calculated spectra and intensities of oscillators one interpreted local bands of optical absorption, luminescence, photoconductivity in pyrolytic boron nitride prior to and subsequent to irradiation by fast neutrons, protons and ions of carbon (50-150 keV). One identified not deep levels of activation of thermally stimulated luminescence and conductivity prior to and subsequent to irradiation

  11. Implications of Graphite Radiation Damage on the Neutronic, Operational, and Safety Aspects of Very High Temperature Reactors

    International Nuclear Information System (INIS)

    Hawari, Ayman I.

    2011-01-01

    In both the prismatic and pebble bed designs of Very High Temperature Reactors (VHTR), the graphite moderator is expected to reach exposure levels of 10 21 to 10 22 n/cm 2 over the lifetime of the reactor. This exposure results in damage to the graphite structure. In this work, molecular dynamic and ab initio molecular static calculations will be used to: (1) simulate radiation damage in graphite under various irradiation and temperature conditions, (2) generate the thermal neutron scattering cross sections for damaged graphite, and (3) examine the resulting microstructure to identify damage formations that may produce the high-temperature Wigner effect. The impact of damage on the neutronic, operational and safety behavior of the reactor will be assessed using reactor physics calculations. In addition, tests will be performed on irradiated graphite samples to search for the high-temperature Wigner effect, and phonon density of states measurements will be conducted to quantify the effect on thermal neutron scattering cross sections using these samples.

  12. Graphite Oxidation Thermodynamics/Reactions

    International Nuclear Information System (INIS)

    Propp, W.A.

    1998-01-01

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

  13. Food irradiation

    International Nuclear Information System (INIS)

    Sato, Tomotaro; Aoki, Shohei

    1976-01-01

    Definition and significance of food irradiation were described. The details of its development and present state were also described. The effect of the irradiation on Irish potatoes, onions, wiener sausages, kamaboko (boiled fish-paste), and mandarin oranges was evaluated; and healthiness of food irradiation was discussed. Studies of the irradiation equipment for Irish potatoes in a large-sized container, and the silo-typed irradiation equipment for rice and wheat were mentioned. Shihoro RI center in Hokkaido which was put to practical use for the irradiation of Irish potatoes was introduced. The state of permission of food irradiation in foreign countries in 1975 was introduced. As a view of the food irradiation in the future, its utilization for the prevention of epidemics due to imported foods was mentioned. (Serizawa, K.)

  14. Microstructure and some properties of boron modified graphite USB-15

    Science.gov (United States)

    Chernikov, V. N.; Alimov, V. Kh.; Gorodetsky, A. E.; Sharapov, V. M.; Zakharov, A. P.; Kurolenkin, E. I.

    1992-09-01

    Boronized graphites, in particular USB-15, have some unique properties which make them applicable in thermonuclear confinement experiments. The optimization of USB-15 manufacture technology demands more detailed knowledge of its properties and, first of all, its microstructure. In this connection microstructure of USB-15 was studied in detail by means of X-ray diffraction, microanalysis, analytical TEM, SEM and other methods, both in original state and after annealing at 2300, 2700 and 3100 K. The erosion resistance to low energy bombardment from deuterium plasma was investigated between 350 and 900 K. In the course of 3 keV D +-ion implantation up to a dose ~ 5 × 10 20 D + m -2 deuterium is mainly trapped in vacancy complexes, which were studied by means of TDS and deuterium profiling using SIMS and RGA. The ratio of CD 4 to D 2 molecules emitted from USB-15 during post-implantation thermal desorption is considerably less than that from other types of graphites, indicating that the recombination desorption of deuterium is greatly enhanced by B in solid solution. Summary is given on property evolution of USB-15 (before and after thermal treatment) as a result of irradiation up to a fast neutron fluence ( E > 0.18 MeV) of 2.5 × 10 25 n m -2 in a temperature range 360-1100 K. Thermal annealing of USB-15 in the range 2200-2400 K is recommended for improvement of its thermal strength and radiation stability.

  15. Gamma irradiator

    International Nuclear Information System (INIS)

    Simonet, G.

    1986-09-01

    Fiability of devices set around reactors depends on material resistance under irradiation noticeably joints, insulators, which belongs to composition of technical, safety or physical incasurement devices. The irradiated fuel elements, during their desactivation in a pool, are an interesting gamma irradiation device to simulate damages created in a nuclear environment. The existing facility at Osiris allows to generate an homogeneous rate dose in an important volume. The control of the element distances to irradiation box allows to control this dose rate [fr

  16. Graphite Fluoride Fiber Composites For Heat Sinking

    Science.gov (United States)

    Hung, Ching-Cheh; Long, Martin; Stahl, Mark

    1989-01-01

    Graphite fluoride fiber/polymer composite materials consist of graphite fluoride fibers in epoxy, polytetrafluoroethylene, or polyimide resin. Combines high electrical resistivity with high thermal conductivity and solves heat-transfer problems of many electrical systems. Commercially available in powder form, for use as dry lubricant or cathode material in lithium batteries. Produced by direct fluorination of graphite powder at temperature of 400 to 650 degree C. Applications include printed-circuit boards for high-density power electronics, insulators for magnetic-field cores like those found in alternators and transformers, substrates for thin-film resistors, and electrical-protection layers in aircraft de-icers.

  17. 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

  18. Status of Chronic Oxidation Studies of Graphite

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-01

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

  19. Stable dispersions of polymer-coated graphitic nanoplatelets

    Science.gov (United States)

    Stankovich, Sasha (Inventor); Nguyen, Sonbinh T. (Inventor); Ruoff, Rodney S. (Inventor)

    2011-01-01

    A method of making a dispersion of reduced graphite oxide nanoplatelets involves providing a dispersion of graphite oxide nanoplatelets and reducing the graphite oxide nanoplatelets in the dispersion in the presence of a reducing agent and a polymer. The reduced graphite oxide nanoplatelets are reduced to an extent to provide a higher C/O ratio than graphite oxide. A stable dispersion having polymer-treated reduced graphite oxide nanoplatelets dispersed in a dispersing medium, such as water or organic liquid is provided. The polymer-treated, reduced graphite oxide nanoplatelets can be distributed in a polymer matrix to provide a composite material.

  20. PRELIMINARY RESULTS OF THE AGC-4 IRRADIATION IN THE ADVANCED TEST REACTOR AND DESIGN OF AGC-5 (HTR16-18469)

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Program will irradiate up to six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments are being irradiated over an approximate eight year period to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Very High Temperature Gas Reactor (VHTR), as well as other future gas reactors. The experiments each consist of a single capsule that contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens are not be subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There are also samples taken of the sweep gas effluent to measure any oxidation or off-gassing of the specimens that may occur during initial start-up of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. The third experiment, AGC-3, started its irradiation in late November 2012 and completed in the April of 2014. AGC-4 is currently being irradiated in the ATR. This paper will briefly discuss the preliminary irradiation results

  1. IRRADIATION DEVICE FOR IRRADIATION TESTING OF COATED PARTICLE FUEL AT HANARO

    Directory of Open Access Journals (Sweden)

    BONG GOO KIM

    2013-12-01

    Full Text Available The Korean Nuclear-Hydrogen Technology Development (NHTD Plan will be performing irradiation testing of coated particle fuel at HANARO to support the development of VHTR in Korea. This testing will be carried out to demonstrate and qualify TRISO-coated particle fuel for use in VHTR. The testing will be irradiated in an inert gas atmosphere without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The irradiation device is being loaded and irradiated into the OR5 hole of the in HANARO core from August 2013. The device will be operated for about 150 effective full-power days at a peak temperature of about 1030°C in BOC (Beginning of Cycle during irradiation testing. After a peak burn-up of about 4 atomic percentage and a peak fast neutron fluence of about 1.7×1021 n/cm2, PIE (Post-Irradiation Examination of the irradiated coated particle fuel will be performed at IMEF (Irradiated Material Examination Facility. This paper reviews the design of test rod and irradiation device for coated particle fuel, and discusses the technical results for irradiation testing at HANARO.

  2. 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.

  3. 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

  4. Synthesis of graphene nanoplatelets from peroxosulfate graphite intercalation compounds

    OpenAIRE

    MELEZHYK A.V.; TKACHEV A.G.

    2014-01-01

    Ultrasonic exfoliation of expanded graphite compound obtained by cold expansion of graphite intercalated with peroxodisulfuric acid was shown to allow the creation of graphene nanoplatelets with thickness of about 5-10 nm. The resulting graphene material contained surface oxide groups. The expanded graphite intercalation compound was exfoliated by ultrasound much easier than thermally expanded graphite. A mechanism for the cleavage of graphite to graphene nanoplatelets is proposed. It include...

  5. Submicron-sized boron carbide particles encapsulated in turbostratic graphite prepared by laser fragmentation in liquid medium.

    Science.gov (United States)

    Ishikawa, Yoshie; Sasaki, Takeshi; Koshizaki, Naoto

    2010-08-01

    Submicron-sized B4C spherical particles were obtained by laser fragmentation of large B4C particles dispersed in ethyl acetate. The irradiated surface of large B4C raw particles was heated and melted by laser energy absorption. B4C droplets were then cooled down, and finally B4C spherical particles were obtained. Moreover, each B4C particle obtained was encapsulated in a graphitic layer that is useful for medical functionalization of particles. Thus, obtained B4C particles encapsulated in graphitic layer may have potential uses in boron neutron capture therapy.

  6. Food irradiation

    International Nuclear Information System (INIS)

    Duchacek, V.

    1989-01-01

    The ranges of doses used for food irradiation and their effect on the processed foods are outlined. The wholesomeness of irradiated foods is discussed. The present food irradiation technology development in the world is described. A review of the irradiated foods permitted for public consumption, the purposes of food irradiaton, the doses used and a review of the commercial-scale food irradiators are tabulated. The history and the present state of food processing in Czechoslovakia are described. (author). 1 fig., 3 tabs., 13 refs

  7. Food irradiation

    International Nuclear Information System (INIS)

    Macklin, M.

    1987-01-01

    The Queensland Government has given its support the establishment of a food irradiation plant in Queensland. The decision to press ahead with a food irradiation plant is astonishing given that there are two independent inquiries being carried out into food irradiation - a Parliamentary Committee inquiry and an inquiry by the Australian Consumers Association, both of which have still to table their Reports. It is fair to assume from the Queensland Government's response to date, therefore, that the Government will proceed with its food irradiation proposals regardless of the outcomes of the various federal inquiries. The reasons for the Australian Democrats' opposition to food irradiation which are also those of concerned citizens are outlined

  8. Optical motion control of maglev graphite.

    Science.gov (United States)

    Kobayashi, Masayuki; Abe, Jiro

    2012-12-26

    Graphite has been known as a typical diamagnetic material and can be levitated in the strong magnetic field. Here we show that the magnetically levitating pyrolytic graphite can be moved in the arbitrary place by simple photoirradiation. It is notable that the optical motion control system described in this paper requires only NdFeB permanent magnets and light source. The optical movement is driven by photothermally induced changes in the magnetic susceptibility of the graphite. Moreover, we demonstrate that light energy can be converted into rotational kinetic energy by means of the photothermal property. We find that the levitating graphite disk rotates at over 200 rpm under the sunlight, making it possible to develop a new class of light energy conversion system.

  9. Large Scale Reduction of Graphite Oxide

    Data.gov (United States)

    National Aeronautics and Space Administration — This project seeks to develop an optical method to reduce graphite oxide into graphene efficiently and in larger formats than currently available. Current reduction...

  10. Graphite Oxide and Aromatic Amines : Size Matters

    NARCIS (Netherlands)

    Spyrou, Konstantinos; Calvaresi, Matteo; Diamanti, Evmorfi A. K.; Tsoufis, Theodoros; Gournis, Dimitrios; Rudolf, Petra; Zerbetto, Francesco

    2015-01-01

    Experimental and theoretical studies are performed in order to illuminate, for first time, the intercalation mechanism of polycyclic aromatic molecules into graphite oxide. Two representative molecules of this family, aniline and naphthalene amine are investigated. After intercalation, aniline

  11. Review: BNL Tokamak graphite blanket design concepts

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  12. Immobilization of Rocky Flats Graphite Fines Residue

    International Nuclear Information System (INIS)

    Rudisill, T.S.

    1999-01-01

    The development of the immobilization process for graphite fines has proceeded through a series of experimental programs. The experimental procedures and results from each series of experiments are discussed in this report

  13. Poly (propylene carbonate)/exfoliated graphite nanocomposites ...

    Indian Academy of Sciences (India)

    propylene carbonate)/exfoliated graphite nanocomposites: selective adsorbent for the extraction and detection of gold(III). Sher Bahadar Khan Hadi M Marwani Jongchul Seo Esraa M Bakhsh Kalsoom Akhtar Dowan Kim Abdullah M Asiri. Volume 38 ...

  14. Seismic research on graphite reactor core

    International Nuclear Information System (INIS)

    Lai Shigang; Sun Libin; Zhang Zhengming

    2013-01-01

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

  15. Energy evaluations, graphite corrosion in Bugey I

    International Nuclear Information System (INIS)

    Brisbois, J.; Fiche, C.

    1967-01-01

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

  16. Measurement of the cleavage energy of graphite.

    Science.gov (United States)

    Wang, Wen; Dai, Shuyang; Li, Xide; Yang, Jiarui; Srolovitz, David J; Zheng, Quanshui

    2015-08-28

    The basal plane cleavage energy (CE) of graphite is a key material parameter for understanding many of the unusual properties of graphite, graphene and carbon nanotubes. Nonetheless, a wide range of values for the CE has been reported and no consensus has yet emerged. Here we report the first direct, accurate experimental measurement of the CE of graphite using a novel method based on the self-retraction phenomenon in graphite. The measured value, 0.37±0.01 J m(-2) for the incommensurate state of bicrystal graphite, is nearly invariant with respect to temperature (22 °C≤T≤198 °C) and bicrystal twist angle, and insensitive to impurities from the atmosphere. The CE for the ideal ABAB graphite stacking, 0.39±0.02 J m(-2), is calculated based on a combination of the measured CE and a theoretical calculation. These experimental measurements are also ideal for use in evaluating the efficacy of competing theoretical approaches.

  17. Laser irradiation of carbon–tungsten materials

    International Nuclear Information System (INIS)

    Marcu, A; Lungu, C P; Ursescu, D; Porosnicu, C; Grigoriu, C; Avotina, L; Kizane, G; Marin, A; Osiceanu, P; Grigorescu, C E A; Demitri, N

    2014-01-01

    Carbon–tungsten layers deposited on graphite by thermionic vacuum arc (TVA) were directly irradiated with a femtosecond terawatt laser. The morphological and structural changes produced in the irradiated area by different numbers of pulses were systematically explored, both along the spots and in their depths. Although micro-Raman and Synchrotron-x-ray diffraction investigations have shown no carbide formation, they have shown the unexpected presence of embedded nano-diamonds in the areas irradiated with high fluencies. Scanning electron microscopy images show a cumulative effect of the laser pulses on the morphology through the ablation process. The micro-Raman spatial mapping signalled an increased percentage of sp 3 carbon bonding in the areas irradiated with laser fluencies around the ablation threshold. In-depth x-ray photoelectron spectroscopy investigations suggested a weak cumulative effect on the percentage increase of the sp 2 -sp 3 transitions with the number of laser pulses just for nanometric layer thicknesses. (paper)

  18. Effect of Graphite Doped TiO2 Nanoparticles on Smoke Degradation

    International Nuclear Information System (INIS)

    Roshasnorlyza Hazan; Mohamad Shahrizal Md Zain; Natrah Syafiqah Rosli

    2016-01-01

    Secondhand smoke affects in the same way as regular smoker. The best solution is to purify the air efficiently and effectively. In this study, we were successfully doped TiO 2 nanoparticle with graphite to accelerate the degradation of cigarette smoke. The graphite doped and undoped TiO 2 nanoparticles were prepared from synthetic rutile using alkaline fusion method and their photo catalytic activity were investigated under visible light irradiation. The photo catalytic activity of the TiO 2 nanoparticles was analyzed in terms of their particle size analysis, crystallization and optical band gap. TiO 2 nanoparticle act as photo catalyzer by utilization of light energy to excite electron-hole pairs in smoke degradation processes. With the aided from graphite in TiO 2 nanoparticles, the smoke degradation was accelerate up to 44.4 %. In this case, graphite helps to reduce optical band gap of TiO 2 nanoparticle, thus increasing excitation of electron from valence band to conduction band. (author)

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

    International Nuclear Information System (INIS)

    McNally, K.; Warren, N.; Fahad, M.; Hall, G.; Marsden, B.J.

    2017-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  1. Natural and synthetic graphite powders: production and main industrial uses; Graphites naturels et synthetiques pulverulents

    Energy Technology Data Exchange (ETDEWEB)

    Wilhelm, H.A. [Timcal Graphite et Carton, CH (Switzerland); L' heureux, J. [Timcal Graphite et Carton, Quebec (Canada)

    2006-03-15

    Large volumes of natural and synthetic graphite powders are yearly used worldwide in applications as different as alkaline and lithium-ion batteries, refractory, lubricant and carbon brushes for instance... After a short description of the conventional processes used to obtain these powders, the role of graphite material into chosen applications is detailed. (authors)

  2. Foodstuff irradiation

    International Nuclear Information System (INIS)

    1982-01-01

    Report written on behalf of the Danish Food Institute summarizes national and international rules and developments within food irradiation technology, chemical changes in irradiated foodstuffs, microbiological and health-related aspects of irradiation and finally technological prospects of this conservation form. Food irradiatin has not been hitherto applied in Denmark. Radiation sources and secondary radiation doses in processed food are characterized. Chemical changes due to irradiation are compared to those due to p.ex. food heating. Toxicological and microbiological tests and their results give no unequivocal answer to the problem whether a foodstuff has been irradiated. The most likely application fields in Denmark are for low radiation dosis inhibition of germination, riping delay and insecticide. Medium dosis (1-10 kGy) can reduce bacteria number while high dosis (10-50 kGy) will enable total elimination of microorganisms and viruses. Food irradiation can be acceptable as technological possibility with reservation, that further studies follow. (EG)

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  4. Graphite matrix materials for nuclear waste isolation

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, W.C.

    1981-06-01

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

  5. 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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  7. Long history of 36Cl assessment of graphite waste by EDF engineering and the latest suggested developments. 36Cl assessment of EDF graphite waste and the latest suggested developments

    International Nuclear Information System (INIS)

    Poncet, B.

    2017-01-01

    About 17.000 tons of irradiated graphite waste will be produced from the decommissioning of the six gas-cooled nuclear reactors operated by Electricite De France (EDF) company. Determining the radionuclide content of this waste is an important legal commitment for both safety reasons and best suited management strategy. As evidenced by numerous studies nuclear graphite is a very complex material that cannot be considered on an analytical viewpoint as any usual homogeneous material. Radionuclide measurements in irradiated graphite exhibit very high discrepancies especially when corresponding to precursors at trace level. This huge discrepancy cannot be avoided and can be easily explained by Pierre Gy's theory of sampling applied to finely divided materials. The assessment of a radionuclide inventory only based on a few number of radiochemical measurements leads in most cases to a gross over- or under-estimation that can be detrimental to graphite waste management. Prior to 2005, the EDF initial version for the radiological inventory of graphite was based on the maximum values of the measures, namely a very pessimistic way. In 2008, a scientific method was developed by EDF to evaluate the inventory by reverse activation calculation, in order to limit the overestimation. The radiological inventory computation principles can be compared to those of all engineering studies. First, the calculation is simplified by taking margins, and if the results are not satisfactory enough, some simplifications are suppressed to reduce these margins even if computation becomes more complicated. On EDF piles, the gain obtained in 2008 represented a factor 50 relative to the initial version of 36 Cl, even with a very penalizing multiplicative factor for uncertainty. Today, the accurate calculation of the uncertainty no longer justifies such a factor. A gain of about 150 now occurs compared to the initial version of 36 Cl. It is more than 2 orders of magnitude, which is already

  8. Hemibody irradiation

    International Nuclear Information System (INIS)

    Schen, B.C.; Mella, O.; Dahl, O.

    1992-01-01

    In a large number of cancer patients, extensive skeletal metastases or myelomatosis induce vast suffering, such as intolerable pain and local complications of neoplastic bone destruction. Analgetic drugs frequently do not yield sufficient palliation. Irradiation of local fields often has to be repeated, because of tumour growth outside previously irradiated volumes. Wide field irradiation of the lower or upper half of the body causes significant relief of pain in most patients. Adequate pretreatment handling of patients, method of irradiation, and follow-up are of importance to reduce side effects, and are described as they are carried out at the Department of Oncology, Haukeland Hospital, Norway. 16 refs., 2 figs

  9. Investigation on Conversion of I-Graphite from Decommissioning of Chernobyl NPP into a Stable Waste Form Acceptable for Long Term Storage and Disposal

    International Nuclear Information System (INIS)

    Zlobenko, Borys; Fedorenko, Yriy; Yatzenko, Victor; Shabalin, Borys; Skripkin, Vadim

    2016-01-01

    For Ukraine, the main radiocarbon ( 14 C) source is irradiated graphite from Chernobyl Nuclear Power Plant. The ChNPP is a decommissioned nuclear power station about 14 km northwest of the city of Chernobyl, and 110 km north of Kyiv. The ChNPP had four RBMK reactor units. The commissioning of the first reactor in 1977 was followed by reactor No. 2 (1978), No. 3 (1981), and No.4 (1983). Reactors No.3 and 4 were second generation units, whereas Nos.1 and 2 were first-generation units. RBMK is an acronym for ''High Power Channel-type Reactor'' of a class of graphite-moderated nuclear power reactor with individual fuel channels that uses ordinary water as its coolant and graphite as its moderator. The combination of graphite moderator and water coolant is found in no other type of nuclear reactor

  10. Gas transport in graphitic materials

    International Nuclear Information System (INIS)

    Hoinkis, E.

    1995-02-01

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

  11. Graphitic matrix materials for spherical HTR fuel elements

    International Nuclear Information System (INIS)

    Rind, W.; Schulze, R.E.; Schulze, H.A.

    1981-12-01

    The present report comprises the essential results of material development and irradiation testing of graphitic matrix materials for spherical HTR fuel elements and completes the documentation of the irradiation data for 20 matrix materials (Juel-1702). The main emphasis is given to the matrices A3-3 (standard matrix) and A3-27 (matrix synthesized resin), both of which are being used as structural materials for the fuel elements of the AVR and the THTR respectively. In addition, comparisons are made between 18 A3-variants and the standard matrix A3-3. It is shown that three of the variants come into question as a potential for use. The results described were obtained in the framework of the HTR project 'Hochtemperaturreaktor-Brennstoffkreislauf' (HBK), in which are involved the Gesellschaft fuer Hochtemperaturreaktor-Technik mbH, Hochtemperaturreaktor-Brennelemente GmbH, Hochtemperatur-Reaktorbau GmbH, Kernforschungsanlage Juelich GmbH, NUKEM GmbH, and Sigri Elektrographit GmbH/Ringsdorff-Werke GmbH. The project is sponsored by the 'Bundesministerium fuer Forschung und Technologie' and by the state of 'Nordrhein-Westfalen'. (orig.) [de

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

    International Nuclear Information System (INIS)

    Li Junfeng

    2016-01-01

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

  13. Study of radiation-induced desorption of gases out of graphite materials

    International Nuclear Information System (INIS)

    Vorob'ev, A.E.; Katrich, N.P.; Sidel'nikova, N.S.; Fedorenko, A.I.

    1987-01-01

    The study results of radiative stimulated desorption of dissolved gases and helium from GMZ and GSP graphites under irradiation by 4 He + ions (E=31 keV) at T=78 K and in backout process after iradiation (T up to 400 K) are presented in paper. It have been studied the initial samples and the samples preliminary irradiated by electrons (E=1.5 MeV) by thermal and fast neutrons and by clusters of hydrogen and helium plasmas (E10 keV). It has been shown that the desorption of impurity pases (CO, C 2 H 4 , N 2 , H 2 ) is minimum for GSP samples and is of order more for GMZ samples. The all types of preliminary irradiation result in the increasing of impurity gases desorption. The activation energies for impurity gases and helium have been determined from thermodesorbtion data

  14. Glassy carbon coated graphite for nuclear applications

    International Nuclear Information System (INIS)

    Delpeux S; Cacciaguerra T; Duclaux L

    2005-01-01

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

  15. Short carbon fiber reinforced electrically conductive aromatic polydisulfide/expanded graphite nanocomposites

    Energy Technology Data Exchange (ETDEWEB)

    Song, L.N. [Guangzhou Institute of Chemistry, Chinese Academy of Sciences, P.O. Box 1122, Guangzhou 510650 (China); Xiao, M. [Institute of Energy and Environmental Materials, School of Physics and Engineering, Sun Yat-Sen University, Guangzhou 510275 (China); Li, X.H. [Institute of Energy and Environmental Materials, School of Physics and Engineering, Sun Yat-Sen University, Guangzhou 510275 (China); Meng, Y.Z. [Institute of Energy and Environmental Materials, School of Physics and Engineering, Sun Yat-Sen University, Guangzhou 510275 (China) and Guangzhou Institute of Chemistry, Chinese Academy of Sciences, P.O. Box 1122, Guangzhou 510650 (China)]. E-mail: stdpmeng@zsu.edu.cn

    2005-09-15

    Expanded graphite (EG) was prepared by the exfoliation of expandable graphite under microwave irradiation. Aromatic polydisulfide/EG nanocomposites were then fabricated by absorbing cyclic (arylene disulfide) oligomers into the pores of EG. Subsequently, the nanocomposite precursor was hot-molded at 200 deg C to carry out simultaneously the in situ ring-opening polymerization of these oligomers via free radical mechanism. The resulting aromatic polydisulfide/EG nanocomposite exhibited a intercalated nanostructure as evidenced by transmission electron microscopy (TEM) observation. Short carbon fiber (SCF) was used to further reinforce aromatic polydisulfide/EG nanocomposites. Consequently, the ternary polydisulfide/EG/SCF nanocomposites showed superior mechanical properties and good electrical conductivity. The ternary nanocomposites can be used as electrically conductive materials to prepare the bipolar plates of polymer electrolyte membrane fuel cell.

  16. The construction of the graphite calorimeter GR9 at LNE-LNHB (geometrical and technical consideration)

    International Nuclear Information System (INIS)

    Ostrowsky, A.; Daures, J.

    2008-01-01

    Calorimetry is the most direct dosimetric technique to reach absorbed dose. A calorimeter gives direct access to the energy imparted to matter by ionizing radiation per mass unit by measuring the heat quantity Q produced under irradiation in its sensitive element which is thermally insulated. Graphite was chosen as construction material because all the energy imparted to graphite by ionizing radiation is converted into heat. Thermistors are used for temperature measurements as well as for the electrical heating of the different bodies of the calorimeter. The construction of a calorimeter is the result of a compromise between dosimetric requirements and mechanical constraints. The difficulties encountered are examined and the solutions chosen are detailed. All technical data are gathered in this document. The aim is to provide a practical operative instruction and guidance document, which can help interested laboratories in designing such an instrument. The electrical and thermal tests have shown a good behaviour of the GR9 calorimeter

  17. AGC-2 Specimen Post Irradiation Data Package Report

    Energy Technology Data Exchange (ETDEWEB)

    Windes, William Enoch [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rohrbaugh, David T. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cottle, David L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    This report documents results of the post-irradiation examination material property testing of the creep, control, and piggyback specimens from the irradiation creep capsule Advanced Graphite Creep (AGC)-2 are reported. This is the second of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades. The AGC-2 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 600°C and to a peak dose of 5 dpa (displacements per atom). One-half of the creep specimens were subjected to mechanical stresses (an applied stress of either 13.8, 17.2, or 20.7 MPa) to induce irradiation creep. All post-irradiation testing and measurement results are reported with the exception of the irradiation mechanical strength testing, which is the last destructive testing stage of the irradiation testing program. Material property tests were conducted on specimens from 15 nuclear graphite grades using a similar loading configuration as the first AGC capsule (AGC-1) to provide easy comparison between the two capsules. However, AGC-2 contained an increased number of specimens (i.e., 487 total specimens irradiated) and replaced specimens of the minor grade 2020 with the newer grade 2114. The data reported include specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and volume data necessary to derive density, elastic constants (Young’s modulus, shear modulus, and Poisson’s ratio) from ultrasonic time-of-flight velocity measurements, Young’s modulus from the fundamental frequency of vibration, electrical resistivity, and thermal diffusivity and thermal expansion data from 100–500°C. No data outliers were determined after all measurements were completed. A brief statistical analysis was performed on the irradiated data and a limited comparison between

  18. Food irradiation

    International Nuclear Information System (INIS)

    Mercader, J.P.; Emily Leong

    1985-01-01

    The paper discusses the need for effective and efficient technologies in improving the food handling system. It defines the basic premises for the development of food handling. The application of food irradiation technology is briefly discussed. The paper points out key considerations for the adoption of food irradiation technology in the ASEAN region (author)

  19. Food irradiation

    International Nuclear Information System (INIS)

    Goodburn, K.E.

    1987-01-01

    The use of food preservation by irradiation over 30 years is reviewed in outline. Sprout inhibition, soft fruit preservation, suppression of salmonella in poultry, radiolytic products in foods, the detection of irradiated foods and safety studies so far are considered. (U.K.)

  20. Irradiation proctitis

    International Nuclear Information System (INIS)

    Minami, Akira

    1977-01-01

    Literatures on late rectal injuries are discussed, referring to two patients with uterine cervical cancer in whom irradiation proctitis occurred after telecobalt irradiation following uterine extirpation. To one patients, a total of 5000 rads was irradiated, dividing into 250 rads at one time, and after 3 months, irradiation with a total of 2000 rads, dividing into 200 rads at one time, was further given. In another one patient, two parallel opposing portal irradiation with a total of 6000 rads was given. About a year after the irradiation, rectal injuries and cystitis, accompanying with hemorrhage, were found in both of the patients. Rectal amputation and proctotoreusis were performed. Cystitis was treated by cystic irradiation in the urological department. Pathohistological studies of the rectal specimen revealed atrophic mucosa, and dilatation of the blood vessels and edema in the colonic submucosa. Incidence of this disease, term when the disease occurs, irradiation dose, type of the disease, treatment and prevention are described on the basis of the literatures. (Kanao, N.)

  1. Food irradiation

    International Nuclear Information System (INIS)

    Matsuyama, Akira

    1990-01-01

    This paper reviews researches, commentaries, and conference and public records of food irradiation, published mainly during the period 1987-1989, focusing on the current conditions of food irradiation that may pose not only scientific or technologic problems but also political issues or consumerism. Approximately 50 kinds of food, although not enough to fill economic benefit, are now permitted for food irradiation in the world. Consumerism is pointed out as the major factor that precludes the feasibility of food irradiation in the world. In the United States, irradiation is feasible only for spices. Food irradiation has already been feasible in France, Hollands, Belgium, and the Soviet Union; has under consideration in the Great Britain, and has been rejected in the West Germany. Although the feasibility of food irradiation is projected to increase gradually in the future, commercial success or failure depends on the final selection of consumers. In this respect, the role of education and public information are stressed. Meat radicidation and recent progress in the method for detecting irradiated food are referred to. (N.K.) 128 refs

  2. Food irradiation

    International Nuclear Information System (INIS)

    Webb, T.; Lang, T.

    1990-01-01

    There is a growing tendency worldwide to apply ionizing radiation for food preservation, or for delaying growth or ripening processes. However, research into the effects of such irradiation with ionizing radiation is lagging behind, leaving the knowledge about effects and possible hazards incomplete. The authors very carefully have analysed the available research results and present a detailed account of the current scientific knowledge and assessments. Their conclusion is: irradiated food is to be considered noxious unless its wholesomeness has been unambiguously proven. Consumers so far had not much chance to raise their voice in the debate about the wholesomeness of food irradiation, or have not been heard, the authors say. They call for establishing a European and a worldwide information network to bring together opponents to and information speaking against food irradiation, in order to create a counterweight to the market strategies of the pro-irradiation industry, and to launch initiatives on the political level. (orig./HP) [de

  3. Properties of graphite composites based on natural and synthetic graphite powders and a phenolic novolac binder

    Science.gov (United States)

    Magampa, P. P.; Manyala, N.; Focke, W. W.

    2013-05-01

    Model graphite composites, similar to those used in nuclear applications as encasement material in fuel pebbles, were prepared by uniaxial cold compression moulding. They contained natural flake graphite, synthetic graphite and 20 wt.% phenolic novolac resin binder. The materials were carbonised at 900 °C in a nitrogen atmosphere and then annealed at 1800 °C in helium atmosphere. The X-ray diffraction studies showed that the graphite in these composites had hexagonal crystal structure after annealing. Raman spectroscopy revealed the presence of the structurally disordered phase derived from the carbonised resin. Optical microscopy revealed a flake-like microstructure for composites containing mainly natural graphite and needle-coke like particles for composites containing mainly synthetic graphite. The composites featured anisotropic property behaviour as the particles were partially aligned in a direction perpendicular to the compression direction. Thermogravimetric analysis studies showed that the annealed graphite composites were stable in air to 650 °C. The linear thermal expansion coefficients measured by thermomechanical analysis (20-600 °C) in the direction of pressing were in the range 5-9 × 10-6 K-1 and in the range 1.2-2 × 10-6 K-1 in the direction normal to pressing. The thermal conductivity of the composites were measured using Xenon flash method from 100 to 1000 °C and the values ranged from 19 to 30 W m-1 K-1.

  4. Carbon Nanotubes Growth on Graphite Fibers

    Science.gov (United States)

    Zhu, Shen; Su, Ching-Hua; Lehoczky, S. L.; Muntele, I.; Ila, D.; Curreri, Peter A. (Technical Monitor)

    2002-01-01

    Carbon nanotubes (CNT) were synthesized on graphite fibers by thermal Chemical Vapor Deposition (CVD). On the fiber surface, iron nanoparticles are coated and act as catalysts for CNT growth. The growth temperature ranges from 550 to 1000 C at an ambient pressure. Methane and hydrogen gases with methane contents of 10% to 100% are used for the CNT synthesis. At high growth temperatures (greater than 800 C), the rapid inter-diffusion of the transition metal iron on the graphite surface results in a rough fiber surface with no CNT grown on the surface. When the growth temperature is relatively low (650 - 800 C), CNT are fabricated on the graphite surface with catalytic particles on the nanotube top ends. Using micro Raman spectroscopy in the breath mode region, single-walled or multi-walled CNT can be determined, depending on methane concentrations.

  5. Graphite moderated reactor for thermoelectric generation

    International Nuclear Information System (INIS)

    Akazawa, Issei; Yamada, Akira; Mizogami, Yorikata

    1998-01-01

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

  6. Beryllium irradiation element analysis in the IEA-R1m reactor; Analise do elemento de irradiacao de berilio no reator IEA-R1m

    Energy Technology Data Exchange (ETDEWEB)

    Ricci Filho, Walter

    1998-07-01

    The IEA-R1 reactor at IPEN-Sao Paulo has undergone a modernization to increase its operating power to 5 MW, in order to allow a more efficient production of the {sup 99} Mo radioisotope. An irradiation element made of Be was acquired for the reactor and studies have been carried out to determine its performance when compared to other irradiators available in the reactor namely, the water and graphite irradiation elements. The results obtained showed some advantages of the Beryllium irradiation element for producing {sup 99} Mo: the epithermal neutron flux in the Be irradiation element is approximately 22% greater than that in the graphite irradiation element and 12% greater than that of the water irradiation element; the neutron reaction rate in molybdenum wires inside in irradiation capsule filled with Mo O{sub 3} 10,6% greater them that in the water irradiation element in the same conditions; the negative reactivity introduced in the reactor by the Be irradiation element it substantially smaller than the those introduced by the other elements: -1636 pcm for the Be irradiator, -2977 for the water irradiator and -2568 pcm for the graphite irradiator. It is possible to conclude that the production of the {sup 99} Mo radioisotope with the Be irradiation element can be increased by 12 to 15% in the IEA-R1m reactor. It also requires less fuel for the reactor operation due to the smaller negative reactivity introduced in the reactor core. (author)

  7. 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)

  8. Graphite crystals grown within electromagnetically levitated metallic droplets

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  9. 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

  10. London forces in highly oriented pyrolytic graphite

    Directory of Open Access Journals (Sweden)

    L.V. Poperenko

    2017-07-01

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

  11. HIGH TEMPERATURE REFRACTORY COATING FOR GRAPHITE MOLDS

    Science.gov (United States)

    Stoddard, S.D.

    1958-10-21

    An improved foundry mold coating for use with graphite molds used in the casting of uranium is presented. The refractory mold coating serves to keep the molten uranium from contact with graphite of the mold and thus prevents carbon pickup by the molten metal. The refractory coating is made by dry mixing certain specific amounts of aluminum oxide, bentonite, Tennessee ball clay, and a soluble silicate salt. Water is then added to the mixture and the suspension thus formed is applied by spraying onto the mold.

  12. Vacuum brazing of graphite-metals

    International Nuclear Information System (INIS)

    Jacquot, P.; Coll, B.; Gabriel, M.; Speri, R.

    1989-01-01

    This conference paper discusses the brazing in vacuum of stainless steel (304 L) and graphite. In order to reduce stresses induced in the brazed system, molybdenum and copper foils are inserted between the two base materials. The filler metal used for brazing is the alloy 69AG27Cu4Ti (Ticusil). The structure of the metal-graphite joint is explained in detail, and a microhardness profile is given. This type of joint is primarily applied in devices for thermonuclear fusion (Tokamak devices). (MM) [de

  13. Cluster Ion Implantation in Graphite and Diamond

    DEFF Research Database (Denmark)

    Popok, Vladimir

    2014-01-01

    Cluster ion beam technique is a versatile tool which can be used for controllable formation of nanosize objects as well as modification and processing of surfaces and shallow layers on an atomic scale. The current paper present an overview and analysis of data obtained on a few sets of graphite...... and diamond samples implanted by keV-energy size-selected cobalt and argon clusters. One of the emphases is put on pinning of metal clusters on graphite with a possibility of following selective etching of graphene layers. The other topic of concern is related to the development of scaling law for cluster...

  14. 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

  15. Large Scale Reduction of Graphite Oxide Project

    Science.gov (United States)

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

    2015-01-01

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

  16. 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

  17. 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.

  18. Thermal Properties of G-348 Graphite

    Energy Technology Data Exchange (ETDEWEB)

    McEligot, Donald [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 College of New York, NY (United States)

    2016-05-01

    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. Data are tabulated and a preliminary correlation for the thermal conductivity is presented as a function of temperature from laboratory temperature to 1000C.

  19. Diamond detectors with laser induced surface graphite electrodes

    Energy Technology Data Exchange (ETDEWEB)

    Komlenok, M. [A.M. Prokorhov General Physics Institute, Russian Academy of Sciences, 38 Vavilova Str., 119991 Moscow (Russian Federation); Bolshakov, A. [A.M. Prokorhov General Physics Institute, Russian Academy of Sciences, 38 Vavilova Str., 119991 Moscow (Russian Federation); Harbin Institute of Technology, 92 Xidazhi Str., 150001 Harbin (China); Ralchenko, V. [A.M. Prokorhov General Physics Institute, Russian Academy of Sciences, 38 Vavilova Str., 119991 Moscow (Russian Federation); Harbin Institute of Technology, 92 Xidazhi Str., 150001 Harbin (China); National Research Nuclear University “MEPhI”, Kashirskoye shosse, 31, 115409 Moscow (Russian Federation); Konov, V. [A.M. Prokorhov General Physics Institute, Russian Academy of Sciences, 38 Vavilova Str., 119991 Moscow (Russian Federation); National Research Nuclear University “MEPhI”, Kashirskoye shosse, 31, 115409 Moscow (Russian Federation); Conte, G. [Department of Sciences, University Roma Tre and INFN, Via Vasca Navale, 84-00148 Rome (Italy); CNR-ISM, Institute for Structure of Matter, National Research Council, Via Salaria km 29, 300, Montelibretti (Italy); Girolami, M. [CNR-ISM, Institute for Structure of Matter, National Research Council, Via Salaria km 29, 300, Montelibretti (Italy); Oliva, P. [University Niccolò Cusano, Via don Carlo Gnocchi, 3-00166 Rome (Italy); Mediterranean Institute of Fundamental Physics ‘MIFP’, Via Appia Nuova, 31-00040 Marino (Rome) (Italy); Salvatori, S. [University Niccolò Cusano, Via don Carlo Gnocchi, 3-00166 Rome (Italy)

    2016-11-21

    We report on the response of metal-less CVD polycrystalline-diamond pixel sensors under β-particles irradiation. A 21×21 array of 0.18×0.18 mm{sup 2} pixels was realized on one side of a 10.0×10.0×0.5 mm{sup 3} polycrystalline diamond substrate by means of laser induced surface graphitization. With the same technique, a large graphite contact, used for detector biasing, was fabricated on the opposite side. A coincidence detecting method was used with two other reference polycrystalline diamond detectors for triggering, instead of commonly used scintillators, positioned in the front and on the back of the sensor-array with respect to the impinging particles trajectory. The collected charge distribution at each pixel was analyzed as a function of the applied bias. No change in the pulse height distribution was recorded by inverting the bias voltage polarity, denoting contacts ohmicity and symmetry. A fairly good pixel response uniformity was obtained: the collected charge most probable value saturates for all the pixels at an electric field strength of about ±0.6 V/μm. Under saturation condition, the average collected charge was equal to =1.64±0.02 fC, implying a charge collection distance of about 285 µm. A similar result, within 2%, was also obtained for 400 MeV electrons at beam test facility at INFN Frascati National Laboratory. Experimental results highlighted that more than 84% of impinging particles involved only one pixel, with no significant observed cross-talk effects.

  20. 3D microscopy of hydrogen and magnetic force on proton implanted microstructures in graphite

    International Nuclear Information System (INIS)

    Reichart, P.; Cluitmans, J.F.J.; Pakes, C.; Orbons, S.; Jamieson, D.N.

    2005-01-01

    We investigated the depth dependence of magnetic signals in proton irradiated graphite using a tilted microspot implantation followed by combined AFM/MFM analysis. This study is motivated by the not yet independently reproduced discovery of ferromagnetism in carbon materials created by proton irradiation. We present results of 3D hydrogen analysis of pristine and irradiated highly oriented pyrolytic graphite (HOPG). These results, previously presented in collaboration with universities in Leipzig and Munich, are summarized here and reveal a hydrogen level in pristine HOPG less than 0.3 at-ppm and that 2.25 MeV implanted hydrogen is located within a peak confined to the end of range with no evidence of diffusion broadening. For implanted microspots, up to 40 at-% of the implanted hydrogen is not detected, providing support for lateral hydrogen diffusion. Up to 10 16 H-atoms/cm 2 are detected in the near-surface region on all samples, which has not yet been considered in possible mechanisms for creation of ferromagnetism. As theoretical models propose that hydrogen could play a major role in carbon ferromagnetism, this result raises the hypothesis for an effect restricted to the surface. Our preliminary data on magnetic force microscopy of tilted implants show a strong magnetic phase shift localized on the beam entrance point only. (author). 14 refs., 5 figs

  1. 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)

  2. 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.

  3. Food irradiation

    International Nuclear Information System (INIS)

    Cormick, C.

    1987-01-01

    The advantages and disadvantages of food irradiation are discussed. Research into food irradiation has been going on for 40 years, yet studies have yet to demonstrate conclusively whether it is safe or harmful. The Australian House of Representatives has established an inquiry into food irradiation conducted by the Standing Committee on Environment and Conservation. It will inquire into the use of ionising radiation for commercial sterilisation, disinfestation, food preservation and other purposes with particular reference to human health and safety, environmental impacts and the adequacy of assessment and regulatory procedures

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

    Czech Academy of Sciences Publication Activity Database

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

    2016-01-01

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

  5. Food irradiation

    International Nuclear Information System (INIS)

    Paganini, M.C.

    1991-06-01

    Food treatment by means of ionizing energy, or irradiation, is an innovative method for its preservation. In order to treat important volumes of food, it is necessary to have industrial irradiation installations. The effect of radiations on food is analyzed in the present special work and a calculus scheme for an Irradiation Plant is proposed, discussing different aspects related to its project and design: ionizing radiation sources, adequate civil work, security and auxiliary systems to the installations, dosimetric methods and financing evaluation methods of the project. Finally, the conceptual design and calculus of an irradiation industrial plant of tubercles is made, based on the actual needs of a specific agricultural zone of our country. (Author) [es

  6. Food irradiation

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    Food preservation by irradiation is one part of Eisenhower's Atoms for Peace program that is enjoying renewed interest. Classified as a food additive by the Food, Drug, and Cosmetic Act of 1958 instead of a processing technique, irradiation lost public acceptance. Experiments have not been done to prove that there are no health hazards from gamma radiation, but there are new pressures to get Food and Drug Administration approval for testing in order to make commercial use of some radioactive wastes. Irradiation causes chemical reactions and nutritional changes, including the destruction of several vitamins, as well as the production of radiolytic products not normally found in food that could have adverse effects. The author concludes that, lacking epidemiological evidence, willing buyers should be able to purchase irradiated food as long as it is properly labeled

  7. Industrial Applications of Graphite Fluoride Fibers

    Science.gov (United States)

    Hung, Ching-Cheh; Kucera, Donald

    1991-01-01

    Based on fluorination technology developed during 1934 to 1959, and the fiber technology developed during the 1970s, a new process was developed to produce graphite fluoride fibers. In the process, pitch based graphitized carbon fibers are at first intercalated and deintercalated several times by bromine and iodine, followed by several cycles of nitrogen heating and fluorination at 350 to 370 C. Electrical, mechanical, and thermal properties of this fiber depend on the fluorination process and the fluorine content of the graphite fluoride product. However, these properties are between those of graphite and those of PTFE (Teflon). Therefore, it is considered to be a semiplastic. The physical properties suggest that this new material may have many new and unexplored applications. For example, it can be a thermally conductive electrical insulator. Its coefficient of thermal expansion (CTE) can be adjusted to match that of silicon, and therefore, it can be a heat sinking printed circuit board which is CTE compatible with silicon. Using these fibers in printed circuit boards may provide improved electrical performance and reliability of the electronics on the board over existing designs. Also, since it releases fluorine at 300 C or higher, it can be used as a material to store fluorine and to conduct fluorination. This application may simplify the fluorination process and reduce the risk of handling fluorine.

  8. Studies on POM/graphite/Ekonol composites

    Indian Academy of Sciences (India)

    Unknown

    behaviour was also investigated by the friction and wear experiment. The worn surface of the composite was studied by SEM technique, and on its basis, the wear mechanism was analysed. Results show that it was possible to prepare POM/graphite/Ekonol composites of high tribology performance and good mechanical.

  9. Studies on POM/graphite/Ekonol composites

    Indian Academy of Sciences (India)

    POM/graphite/Ekonol composites were prepared by the Torque Rheometer mixing and compression molding, and their hardness, compressive and impact strengths have been tested. The tribology behaviour was also investigated by the friction and wear experiment. The worn surface of the composite was studied by SEM ...

  10. US graphite reactor D&D experience

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  11. Thermoexpanded graphite modification by titanium dioxide

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  12. Ultrafast Multiphoton Thermionic Photoemission from Graphite

    Directory of Open Access Journals (Sweden)

    Shijing Tan

    2017-01-01

    Full Text Available Electronic heating of cold crystal lattices in nonlinear multiphoton excitation can transiently alter their physical and chemical properties. In metals where free electron densities are high and the relative fraction of photoexcited hot electrons is low, the effects are small, but in semimetals, where the free electron densities are low and the photoexcited densities can overwhelm them, the intense femtosecond laser excitation can induce profound changes. In semimetal graphite and its derivatives, strong optical absorption, weak screening of the Coulomb potential, and high cohesive energy enable extreme hot electron generation and thermalization to be realized under femtosecond laser excitation. We investigate the nonlinear interactions within a hot electron gas in graphite through multiphoton-induced thermionic emission. Unlike the conventional photoelectric effect, within about 25 fs, the memory of the excitation process, where resonant dipole transitions absorb up to eight quanta of light, is erased to produce statistical Boltzmann electron distributions with temperatures exceeding 5000 K; this ultrafast electronic heating causes thermionic emission to occur from the interlayer band of graphite. The nearly instantaneous thermalization of the photoexcited carriers through Coulomb scattering to extreme electronic temperatures characterized by separate electron and hole chemical potentials can enhance hot electron surface femtochemistry, photovoltaic energy conversion, and incandescence, and drive graphite-to-diamond electronic phase transition.

  13. Fruit irradiation

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    Food spoilage is a common problem when marketing agricultural products. Promising results have already been obtained on a number of food irradiating applications. A process is described in this paper where irradiation of sub-tropical fruits, especially mangoes and papayas, combined with conventional heat treatment results in effective insect and fungal control, delays ripening and greatly improves the quality of fruit at both export and internal markets

  14. Blood irradiation

    International Nuclear Information System (INIS)

    Chandy, Mammen

    1998-01-01

    Viable lymphocytes are present in blood and cellular blood components used for transfusion. If the patient who receives a blood transfusion is immunocompetent these lymphocytes are destroyed immediately. However if the patient is immunodefficient or immunosuppressed the transfused lymphocytes survive, recognize the recipient as foreign and react producing a devastating and most often fatal syndrome of transfusion graft versus host disease [T-GVHD]. Even immunocompetent individuals can develop T-GVHD if the donor is a first degree relative since like the Trojan horse the transfused lymphocytes escape detection by the recipient's immune system, multiply and attack recipient tissues. T-GVHD can be prevented by irradiating the blood and different centers use doses ranging from 1.5 to 4.5 Gy. All transfusions where the donor is a first degree relative and transfusions to neonates, immunosuppressed patients and bone marrow transplant recipients need to be irradiated. Commercial irradiators specifically designed for irradiation of blood and cellular blood components are available: however they are expensive. India needs to have blood irradiation facilities available in all large tertiary institutions where immunosuppressed patients are treated. The Atomic Energy Commission of India needs to develop a blood irradiator which meets international standards for use in tertiary medical institutions in the country. (author)

  15. Tissue irradiator

    International Nuclear Information System (INIS)

    Hungate, F.P.; Riemath, W.F.; Bunnell, L.R.

    1975-01-01

    A tissue irradiator is provided for the in-vivo irradiation of body tissue. The irradiator comprises a radiation source material contained and completely encapsulated within vitreous carbon. An embodiment for use as an in-vivo blood irradiator comprises a cylindrical body having an axial bore therethrough. A radioisotope is contained within a first portion of vitreous carbon cylindrically surrounding the axial bore, and a containment portion of vitreous carbon surrounds the radioisotope containing portion, the two portions of vitreous carbon being integrally formed as a single unit. Connecting means are provided at each end of the cylindrical body to permit connections to blood-carrying vessels and to provide for passage of blood through the bore. In a preferred embodiment, the radioisotope is thulium-170 which is present in the irradiator in the form of thulium oxide. A method of producing the preferred blood irradiator is also provided, whereby nonradioactive thulium-169 is dispersed within a polyfurfuryl alcohol resin which is carbonized and fired to form the integral vitreous carbon body and the device is activated by neutron bombardment of the thulium-169 to produce the beta-emitting thulium-170

  16. Food irradiation

    International Nuclear Information System (INIS)

    1991-01-01

    Processing of food with low levels of radiation has the potential to contribute to reducing both spoilage of food during storage - a particular problem in developing countries - and the high incidence of food-borne disease currently seen in all countries. Approval has been granted for the treatment of more than 30 products with radiation in over 30 countries but, in general, governments have been slow to authorize the use of this new technique. One reason for this slowness is a lack of understanding of what food irradiation entails. This book aims to increase understanding by providing information on the process of food irradiation in simple, non-technical language. It describes the effects that irradiation has on food, and the plant and equipment that are necessary to carry it out safely. The legislation and control mechanisms required to ensure the safety of food irradiation facilities are also discussed. Education is seen as the key to gaining the confidence of the consumers in the safety of irradiated food, and to promoting understanding of the benefits that irradiation can provide. (orig.) With 4 figs., 1 tab [de

  17. Food irradiation

    International Nuclear Information System (INIS)

    Migdal, W.

    1995-01-01

    A worldwide standard on food irradiation was adopted in 1983 by codex Alimentarius Commission of the Joint Food Standard Programme of the Food and Agriculture Organization (FAO) of the United Nations and The World Health Organization (WHO). As a result, 41 countries have approved the use of irradiation for treating one or more food items and the number is increasing. Generally, irradiation is used to: food loses, food spoilage, disinfestation, safety and hygiene. The number of countries which use irradiation for processing food for commercial purposes has been increasing steadily from 19 in 1987 to 33 today. In the frames of the national programme on the application of irradiation for food preservation and hygienization an experimental plant for electron beam processing has been established in Inst. of Nuclear Chemistry and Technology. The plant is equipped with a small research accelerator Pilot (19 MeV, 1 kW) and industrial unit Electronika (10 MeV, 10 kW). On the basis of the research there were performed at different scientific institutions in Poland, health authorities have issued permissions for irradiation for; spices, garlic, onions, mushrooms, potatoes, dry mushrooms and vegetables. (author)

  18. Spatio-temporal mapping of ablated species in ultrafast laser-produced graphite plasmas

    International Nuclear Information System (INIS)

    Al-Shboul, K. F.; Harilal, S. S.; Hassanein, A.

    2012-01-01

    We studied the spatial and temporal distributions of ionic, neutral, and molecular species generated by femtosecond laser produced plasma under varying ambient nitrogen gas pressures. Plasmas were generated by irradiating planar graphite targets using 40 fs pulses of 800 nm radiation from a Ti:Sapphire laser. The results show that in the presence of an ambient gas, the molecular species spatial extension and lifetime are directly correlated to the evolution of excited ions. The present studies also provide valuable insights into the evolution history of various species and their excitation during ultrafast laser ablation.

  19. Mass spectrometric study of carbon cluster formation in laser ablation of graphite at 355 nm

    CERN Document Server

    Koo, Y M; Lee, K H; Jung, K W

    2002-01-01

    The ablation dynamics and cluster formation of C sub n sup + ions ejected from 355 nm laser ablation of a graphite target in vacuum are investigated using a reflectron time-of-flight (RTOF) mass spectrometer. At low laser fluence, odd-numbered cluster ions with 3 =30) are produced at relatively long delay times, indicating that atoms or small carbon clusters aggregate during plume propagation. The dependence of the intensity of ablated C sub n sup + ions on delay time after laser irradiation shows that the most probable velocity of each cluster ion decreases with cluster size.

  20. Radiations from fuel channels in a graphite-carbon dioxide gas reactor

    International Nuclear Information System (INIS)

    Devillers, Christian; Duco, Jacques; Lafore, Pierre; Le Dieu De Ville, Alain; Sonnet, Alain

    1964-10-01

    The authors report the study of the flow of equivalent thermal neutrons going out of fuel channels of a graphite-gas reactor (the importance of this flow is due to radiative captures in the iron of metallic structures of the enclosure above the core or in the enclosure concrete). They also address the gamma irradiation from the direct view of fuel elements, channel walls at the core level, and reflector, and gamma rays emitted by radiative capture in core weight plate walls. Calculation methods are proposed for the case of a reactor with or without fake cartridge. They compare flows and doses measured on an EDF reactor and those calculated

  1. Smooth Particle Hydrodynamics Simulation of Micro-Cup-Extrusion Using a Graphit-ic Coating

    Directory of Open Access Journals (Sweden)

    Li Shi-Cheng

    2014-01-01

    Full Text Available Microextrusion is becoming increasingly important for the manufacturing of microcomponents. However, this reduction in scale to a microlevel means that the influence of friction and the need for suitable lubrication are greatly increased. This study therefore looks at the use of a low-friction and highly wear resistant Graphit-ic coating on the mold-forming section of a microextrusion mold, this coating being applied by a closed-field unbalanced magnetron sputter ion plating technique. A microcup of CuZn33 brass alloy was then extruded, with a wall thickness of 0.45 mm, outside diameter of 2.9 mm, and an internal diameter of 2 mm. The experimental results in which extrusion uses the mold coating with Graphit-ic film are compared against the experimental results in which extrusion uses the mold uncoating with Graphit-ic film. This showed that the load was decreased a lot and the self-lubricating solid coating facilitates a smooth extrusion process. As the extrusion rate was quite high, smoothed particle hydrodynamics method simulations of the extrusion process were conducted, these being then compared with the experimental results. These result showed that the SPH simulation can be applied to show the deformation of materials and predict the load trend.

  2. 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).

  3. Neutronographic investigations into homogeneity and crystalline anisotropy of graphite materials

    International Nuclear Information System (INIS)

    Kajzar, F.; Oles, A.; Pawpowski, K.; Szudek, M.

    1976-01-01

    A numerical methods is proposed for evaluating the share of components having various graphitization degree in the graphite material. This method consists in adjusting the diffraction lines of separated components characterized by a different graphitization degree to the profile of an experimental diffraction line. Results are also given which were obtained by this method for some selected graphite materials manufactured at the Coal Electrodes Factory in Nowy Sacz. Using the technique of neutron diffraction, investigation were carried out in like manner into crystalline anisotropy in the semi-conducting graphite and in the connecting material. (author)

  4. Graphite Oxidation Simulation in HTR Accident Conditions

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed

    2012-10-19

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

  5. 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. PMID:26575374

  6. A novel route to graphite-like carbon supporting SnO{sub 2} with high electron transfer and photocatalytic activity

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Xianjie; Liu, Fenglin; Liu, Bing [Hubei Collaborative Innovation Center for Advanced Organochemical Materials, Hubei University, Wuhan 430062 (China); Ministry of Education Key Laboratory for the Synthesis and Applications of Organic Functional Molecules, Hubei University, Wuhan 430062 (China); Tian, Lihong, E-mail: tian7978@hubu.edu.cn [Hubei Collaborative Innovation Center for Advanced Organochemical Materials, Hubei University, Wuhan 430062 (China); Ministry of Education Key Laboratory for the Synthesis and Applications of Organic Functional Molecules, Hubei University, Wuhan 430062 (China); Hu, Wei; Xia, Qinghua [Hubei Collaborative Innovation Center for Advanced Organochemical Materials, Hubei University, Wuhan 430062 (China); Ministry of Education Key Laboratory for the Synthesis and Applications of Organic Functional Molecules, Hubei University, Wuhan 430062 (China)

    2015-04-28

    Highlights: • Mesoporous nanocomposites that graphite-like carbon supporting SnO{sub 2} are prepared by solvothermal method combined with a post- calcination. • The polyvinylpyrrolidone not only promotes the nucleation and crystallization but also provides the carbon source in the process. • The graphite-like carbon hinders the recombination of photogenerated electron and holes efficiently. • The mesoporous carbon–SnO{sub 2} nanocomposite shows high photocatalytic activity on the degradation of Rhodamine B and glyphosate under simulated sunlight. - Abstract: Mesoporous graphite-like carbon supporting SnO{sub 2} (carbon–SnO{sub 2}) nanocomposites were prepared by a modified solvothermal method combined with a post-calcination at 500 °C under a nitrogen atmosphere. The polyvinylpyrrolidone not only promotes the nucleation and crystallization, but also provides the carbon source in the process. The results of scanning electron microscopy and transmission electron microscopy show a uniform distribution of SnO{sub 2} nanoparticles on the graphite- like carbon surface. Raman and X-ray photoelectron spectra indicate the presence of strong C–Sn interaction between SnO{sub 2} and graphite-like carbon. Photoelectrochemical measurements confirm that the effective separation of electron–hole pairs on the carbon–SnO{sub 2} nanocomposite leads to a high photocatalytic activity on the degradation of Rhodamine B and glyphosate under simulated sunlight irradiation. The nanocomposite materials show a potential application in dealing with the environmental and industrial contaminants under sunlight irradiation.

  7. Calibration of thin-film dosimeters irradiated with 80-120 kev electrons

    DEFF Research Database (Denmark)

    Helt-Hansen, J.; Miller, A.; McEwen, M.

    2004-01-01

    A method for calibration of thin-film dosimeters irradiated with 80-120keV electrons has been developed. The method is based on measurement of dose with a totally absorbing graphite calorimeter, and conversion of dose in the graphite calorimeter to dose in the film dosimeter by Monte Carlo...... calculations. A thermal model was developed to estimate the temperature contributions from the air above the calorimeter that is heated by the electron beam. As an example, Riso B3 thin-film dosimeters were calibrated by 80-120 keV electron irradiation and compared with a calibration carried out at 10 Me...

  8. GRAPHITIZATION OF METASEDIMENTARY ROCKS IN THE WESTERN KONYA

    Directory of Open Access Journals (Sweden)

    Hüseyin KURT

    2000-01-01

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

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

    Science.gov (United States)

    Hoshino, Shuhei; Kawahara, Kazuma; Takeuchi, Nozomi

    2018-01-01

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

  10. Analysis of Natural Graphite, Synthetic Graphite, and Thermosetting Resin Candidates for Use in Fuel Compact Matrix

    Energy Technology Data Exchange (ETDEWEB)

    Trammell, Michael P [ORNL; Pappano, Peter J [ORNL

    2011-09-01

    The AGR-1 and AGR-2 compacting process involved overcoating TRISO particles and compacting them in a steel die. The overcoating step is the process of applying matrix to the OPyC layer of TRISO particles in a rotating drum in order to build up an overcoat layer of desired thickness. The matrix used in overcoating is a mixture of natural graphite, synthetic graphite, and thermosetting resin in the ratio, by weight, of 64:16:20. A wet mixing process was used for AGR-1 and AGR-2, in that the graphites and resin were mixed in the presence of ethyl alcohol. The goal of the wet mixing process was to 'resinate' the graphite particles, or coat each individual graphite particle with a thin layer of resin. This matrix production process was similar to the German, Chinese, Japanese, and South African methods, which also use various amount of solvent during mixing. See Appendix 1 for information on these countries matrix production techniques. The resin used for AGR-1 and AGR-2 was provided by Hexion, specifically Hexion grade Durite SC1008. Durite SC1008 is a solvated (liquid) resole phenolic resin. A resole resin does not typically have a hardening agent added. The major constituent of SC1008 is phenol, with minor amounts of formaldehyde. Durite SC1008 is high viscosity, so additional ethyl alcohol was added during matrix production in order to reduce its viscosity and enhance graphite particle resination. The current compacting scale up plan departs from a wet mixing process. The matrix production method specified in the scale up plan is a co-grinding jet mill process where powdered phenolic resin and graphite are all fed into a jet mill at the same time. Because of the change in matrix production style, SC1008 cannot be used in the jet milling process because it is a liquid. The jet milling/mixing process requires that a suite of solid or powdered resins be investigated. The synthetic graphite used in AGR-1 and AGR-2 was provided by SGL Carbon, grade KRB2000. KRB

  11. The relationship between irradiation induced dimensional change and the coefficient of thermal expansion: a modified Simmons relationship

    International Nuclear Information System (INIS)

    Hall, G.; Marsden, B.J.; Fok, S.L.; Smart, J.

    2003-01-01

    In the 1960s, a theoretical relationship between the dimensional changes and the coefficient of thermal expansion of irradiated graphite was derived by J.H.W. Simmons. The theory was shown to be comparable with experimental observations at low irradiation doses, but shown to diverge at higher irradiation doses. However, various modified versions of this theory have been used as the foundation of design and life prediction calculations for graphite-moderated reactors. This paper re-examines the Simmons relationship, summarising its derivation and assumptions. The relationship was then modified to incorporate the high dose, high strain changes that were assumed to be represented in the changes in Young's modulus with irradiation dose. By scrutinising the behaviour of finite element analyses, it was possible to use a modified Simmons relationship to predict the dimensional changes of an isotropic and anisotropic graphite to high irradiation doses. These issues are important to present high-temperature reactors (HTRs) as the life of HTR graphite components is dependent upon their dimensional change behaviour. A greater understanding of this behaviour will help in the selection and development of graphite materials

  12. Methods used to address protection issues in graphite-gas reactors

    International Nuclear Information System (INIS)

    Brisbois, Jacques

    1969-01-01

    Protection issues are problems related to the presence of neutron and gamma radiations: biological dose, material heating, Wigner effect, gas ionization. The objective is then to determine the biological doses which are predictable in some reactor areas, and to assess the effects of irradiation on materials during the reactor lifetime. For these purposes, problems of particle propagation are addressed, and the effects of radiations related to one phenomenon or another are studied with respect to radiation energy. The author first describes methods used to compute general propagation. These are mainly codes using Monte Carlo or multi-group scattering methods. These methods are experimentally controlled on a sub-critical graphite-natural uranium set. In a second part, the author presents the different radiation sources which are to be taken into account in calculations. Then he indicates the nuclear constants to be used in the different codes, as well as the response functions to be used to calculate a specific phenomenon (steel, graphite Wigner effect, so on) from the particle spectrum. In the fourth part, the author describes the different methods which can be used to solve problems which are specific to graphite reactors. In the last part, obtained calculated results are compared with experimental measurements performed on power reactors [fr

  13. Cobalt-doped graphitic carbon nitride photocatalysts with high activity for hydrogen evolution

    Science.gov (United States)

    Chen, Pei-Wen; Li, Kui; Yu, Yu-Xiang; Zhang, Wei-De

    2017-01-01

    Cobalt-doped graphitic carbon nitride (Cosbnd CN) was synthesized by one-step thermal polycondensation using cobalt phthalocyanine (CoPc) and melamine as precursors. The π-π interaction between melamine and CoPc promotes cobalt doping into the framework of g-C3N4. The prepared samples were carefully characterized and the results demonstrated that Co-doped graphitic carbon nitride inhibited the crystal growth of graphitic carbon nitride (CN), leading to larger specific surface area (33.1 m2 g-1) and abundant Co-Nx active sites, narrower band gap energy and more efficient separation of photogenerated electrons and holes. 0.46% Cosbnd CN exhibited higher hydrogen evolution rate (28.0 μmol h-1) under visible light irradiation, which is about 3.0 times of that over the pure CN and about 2.2 times of that over cobalt-doped CN using CoCl2 • 6H2O as a cobalt source. This study provides a valuable strategy to modify CN with enhanced photocatalytic performance.

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

    International Nuclear Information System (INIS)

    Mitchell, M.N.

    2005-01-01

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

  15. A study of rates of (n, f), (n, γ), and (n, 2n) reactions in natU and 232Th produced by the neutron fluence in the graphite set-up (gamma-3) irradiated by 2.33 GeV deuteron beam

    International Nuclear Information System (INIS)

    Adam, J.; Chitra Bhatia; Katovskij, K.

    2011-01-01

    Spallation neutrons produced in a collision of 2.33 GeV deuteron beam with the large lead target are moderated by the thick graphite block surrounding the target and used to activate the radioactive samples of nat U and Th put at the three different positions, identified as holes 'a', 'b' and 'c' in the graphite block. Rates of the (n, f), (n, γ), and (n, 2n) reactions in the two samples are determined using the gamma spectrometry. Ratio of the experimental reaction rates, R(n, 2n)/R(n, f) for the 232 Th and nat U are estimated in order to understand the role of reactions of (n, xn) type in Accelerator Driven Subcritical Systems. For the Th-sample, the ratio is ∼ 54(10)% in case of hole 'a' and ∼ 95(57)% in case of hole 'b' compared to 1.73(20)% for the hole 'a' and 0.710(9)% for the hole 'b' in case of the nat U sample. Also the ratio of fission rates in uranium to thorium, nat U(n, f)/ 232 Th(n, f), is ∼ 11.2(17) in case of hole 'a' and 26.8(85) in hole 'b'. Similarly, ratio 238 U(n, 2n)/ 232 Th(n, 2n) is 0.36(4) for the hole 'a' and 0.20(10) for the hole 'b' showing that 232 Th is more prone to the (n, xn) reaction than 238 U. All the experimental reaction rates are compared with the simulated ones by generating neutron fluxes at the three holes from MCNPX 2.6c and making use of LA150 library of cross sections. The experimental and calculated rates of all the three reactions are in good agreement. The transmutation power of the set-up is estimated using the rates of (n, γ) and (n, 2n) reactions for both the samples in the three holes and compared with some of the results of the 'Energy plus Transmutation' set-up and TARC experiment

  16. Gap Size Uncertainty Quantification in Advanced Gas Reactor TRISO Fuel Irradiation Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh T.; Einerson, Jeffrey J.; Hawkes, Grant L.; Lybeck, Nancy J.; Petti, David A.

    2016-10-01

    The Advanced Gas Reactor (AGR)-3/4 experiment is the combination of the third and fourth tests conducted within the tristructural isotropic fuel development and qualification research program. The AGR-3/4 test consists of twelve independent capsules containing a fuel stack in the center surrounded by three graphite cylinders and shrouded by a stainless steel shell. This capsule design enables temperature control of both the fuel and the graphite rings by varying the neon/helium gas mixture flowing through the four resulting gaps. Knowledge of fuel and graphite temperatures is crucial for establishing the functional relationship between fission product release and irradiation thermal conditions. These temperatures are predicted for each capsule using the commercial finite-element heat transfer code ABAQUS. Uncertainty quantification reveals that the gap size uncertainties are among the dominant factors contributing to predicted temperature uncertainty due to high input sensitivity and uncertainty. Gap size uncertainty originates from the fact that all gap sizes vary with time due to dimensional changes of the fuel compacts and three graphite rings caused by extended exposure to high temperatures and fast neutron irradiation. Gap sizes are estimated using as-fabricated dimensional measurements at the start of irradiation and post irradiation examination dimensional measurements at the end of irradiation. Uncertainties in these measurements provide a basis for quantifying gap size uncertainty. However, lack of gap size measurements during irradiation and lack of knowledge about the dimension change rates lead to gap size modeling assumptions, which could increase gap size uncertainty. In addition, the dimensional measurements are performed at room temperature, and must be corrected to account for thermal expansion of the materials at high irradiation temperatures. Uncertainty in the thermal expansion coefficients for the graphite materials used in the AGR-3/4 capsules

  17. Immobilization of carbon-14 from reactor graphite waste by use of self-sustaining reaction in the C-Al-TiO2 system

    International Nuclear Information System (INIS)

    Karlina, O.K.; Klimov, V.L.; Ojovan, M.I.; Pavlova, G.Yu.; Dmitriev, S.A.; Yurchenko, A.Yu.

    2005-01-01

    As a result of long-term neutron irradiation, the long-lived 14 C is produced in the reactor graphite. The exothermic self-sustaining reaction 3C(graphite) + 4Al + 3TiO 2 = 3TiC + 2Al 2 O 3 was proposed for processing of such waste. In doing so, the carbon, including the 14 C, is chemically bound in the stable TiC. The reaction products in the C-Al-TiO 2 system were investigated both by thermodynamic simulation and experimentally in the course of this work

  18. Food irradiation

    International Nuclear Information System (INIS)

    Beishon, J.

    1991-01-01

    Food irradiation has been the subject of concern and controversy for many years. The advantages of food irradiation include the reduction or elimination of dangerous bacterial organisms, the control of pests and insects which destroy certain foods, the extension of the shelf-life of many products, for example fruit, and its ability to treat products such as seafood which may be eaten raw. It can also replace existing methods of treatment which are believed to have hazardous side-effects. However, after examining the evidence produced by the proponents of food irradiation, the author questions whether it has any major contribution to make to the problems of foodborne diseases or world food shortages. More acceptable solutions, he suggests, may be found in educating food handlers to ensure that hygienic conditions prevail in the production, storage and serving of food. (author)

  19. Irradiance gradients

    International Nuclear Information System (INIS)

    Ward, G.J.; Heckbert, P.S.; Technische Hogeschool Delft

    1992-04-01

    A new method for improving the accuracy of a diffuse interreflection calculation is introduced in a ray tracing context. The information from a hemispherical sampling of the luminous environment is interpreted in a new way to predict the change in irradiance as a function of position and surface orientation. The additional computation involved is modest and the benefit is substantial. An improved interpolation of irradiance resulting from the gradient calculation produces smoother, more accurate renderings. This result is achieved through better utilization of ray samples rather than additional samples or alternate sampling strategies. Thus, the technique is applicable to a variety of global illumination algorithms that use hemicubes or Monte Carlo sampling techniques

  20. Sicral F1 graphite-core fuel element behavior in power reactors

    International Nuclear Information System (INIS)

    Rendu, M.

    1987-02-01

    Over 500 000 Sicral F1 graphite-core fuel elements have been manufactured by COGEMA to date and irradiated in GCR power reactors. Since 1963, this type of fuel element has been thoroughly investigated in design studies, in-core and out-of-core tests and post-mortem examinations. This report reviews the current state of knowledge on the irradiation behavior of the components under normal operating conditions and in incident situations (e.g. clad failure). It discusses how this work has led to optimization of the thermal, mechanical metallurgical and neutronic performance in order to obtain a can failure probability of less than 1.6 x 10 -5 , and defines general operating procedures for reactor implementation of this type of fuel element [fr

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  2. Theoretical analysis of the graphitization of a nanodiamond

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-09-26

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

  3. Theoretical analysis of the graphitization of a nanodiamond

    International Nuclear Information System (INIS)

    Kwon, S Joon; Park, Jae-Gwan

    2007-01-01

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

  4. Graphite moderated {sup 252}Cf source

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-08-15

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

  5. Fracture toughness of reactor grade graphites, 3

    International Nuclear Information System (INIS)

    Sato, Sennosuke; Awaji, Hideo; Akuzawa, Hironobu; Kon, Junichi.

    1979-01-01

    In our recent papers, we presented a new technique for determining the thermal shock fracture toughness, using a disk specimen with an edge crack. The thermal shock fracture toughness is defined as K sub( ic)k/Eα(K sub( ic) standing for fracture toughness; k for thermal conductivity; E for Young's modulus; α for thermal expansion coefficient) and it can be determined en bloc by measuring the threshold electric power of the arc discharge heating produced when an edge crack propagates in the disk. The value obtained is the fracture toughness corresponding to the thermal shock resistance defined as σk/Eα (σ standing for tensile strength). The experimental data shown in the following discussion concern themselves with four kinds of reactor grade graphite and some varieties of electrode graphite. (author)

  6. 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%

  7. Atomic resolution images of graphite in air

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-12-01

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

  8. Electron oxidation of graphite by fluorospecies

    Energy Technology Data Exchange (ETDEWEB)

    Rosenthal, G.L.

    1984-09-01

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

  9. Temperature dependence of phonons in pyrolitic graphite

    International Nuclear Information System (INIS)

    Brockhouse, B.N.; Shirane, G.

    1977-01-01

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

  10. Electronic chip cooling system using graphite fins

    OpenAIRE

    Xue , Dong; Wu , Long; Xun , Lian

    2017-01-01

    International audience; As electronic devices get smaller, cooling systems with higher thermal efficiency is demanding by fast growing electronic industry. Great amount of research has been performed on the cooling systems but research on the materials of the cooling systems needs more work. Graphite with high thermal conductivity and light weight is a great candidate to be used in electronic devices. The bottleneck of using graphene in the cooling systems is the thermal transport among the i...

  11. Cobalt-doped graphitic carbon nitride photocatalysts with high activity for hydrogen evolution

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Pei-Wen; Li, Kui; Yu, Yu-Xiang; Zhang, Wei-De, E-mail: zhangwd@scut.edu.cn

    2017-01-15

    Graphical abstract: Cobalt-doped graphitic carbon nitride (Co−CN) was synthesized by one-step thermal polycondensation using cobalt phthalocyanine and melamine as precursors. The obtained photocatalysts display high and stable activity for photocatalytic generation of hydrogen through water splitting. - Highlights: • Cobalt-doped g-C{sub 3}N{sub 4} photocatalysts were prepared. • High and stable visible light photocatalytic activity for H{sub 2} evolution. • Efficient separation and transfer of photo-induced electron-hole pairs. - Abstract: Cobalt-doped graphitic carbon nitride (Co−CN) was synthesized by one-step thermal polycondensation using cobalt phthalocyanine (CoPc) and melamine as precursors. The π-π interaction between melamine and CoPc promotes cobalt doping into the framework of g-C{sub 3}N{sub 4}. The prepared samples were carefully characterized and the results demonstrated that Co-doped graphitic carbon nitride inhibited the crystal growth of graphitic carbon nitride (CN), leading to larger specific surface area (33.1 m{sup 2} g{sup −1}) and abundant Co-N{sub x} active sites, narrower band gap energy and more efficient separation of photogenerated electrons and holes. 0.46% Co−CN exhibited higher hydrogen evolution rate (28.0 μmol h{sup −1}) under visible light irradiation, which is about 3.0 times of that over the pure CN and about 2.2 times of that over cobalt-doped CN using CoCl{sub 2} ∙ 6H{sub 2}O as a cobalt source. This study provides a valuable strategy to modify CN with enhanced photocatalytic performance.

  12. 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.)

  13. Microwave Irradiation

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 5; Issue 3. Microwave Irradiation - Way to Eco-friendly, Green Chemistry. Rashmi Sanghi. General Article Volume 5 Issue 3 March 2000 pp 77-81. Fulltext. Click here to view fulltext PDF. Permanent link:

  14. Gonadal Irradiation

    African Journals Online (AJOL)

    division either by mitosis or meiosis show the least degree of radiation tolerance. This can be readily demonstrated histologically. Two hours after irradiation the spermatogenetic cell column starts to shrink because of a reduction in the number of spermatogonia, some of which are undergoing abnormal mitosis. Four days ...

  15. ion irradiation

    Indian Academy of Sciences (India)

    Swift heavy ions interact predominantly through inelastic scattering while traversing any polymer medium and produce excited/ionized atoms. Here samples of the polycarbonate Makrofol of approximate thickness 20 m, spin coated on GaAs substrate were irradiated with 50 MeV Li ion (+3 charge state). Build-in ...

  16. Microwave Irradiation

    Indian Academy of Sciences (India)

    Microwave Irradiation. Way to Eco-friendly, Green Chemistry. Rashmi Sanghi. This article highlights with examples, the usefulness of microwaves for carrying out a'variety of organic transfor- mations. Introduction and Background. The rapid heating of food in the kitchen using microwave ovens prompted a number of ...

  17. Preparation of porous carbon films from polyacrylonitrile by proton irradiation and carbonization

    Science.gov (United States)

    Lee, Hwa Su; Baek, Ga Young; Hwang, In-Tae; Jung, Chan-Hee; Choi, Jae-Hak

    2017-12-01

    Porous carbon films (PCFs) are widely used in various application fields, such as separation and water desalination as well as energy storage and conversion. In this study, PCFs were prepared from polyacrylonitrile (PAN) by proton irradiation and carbonization. Porous PAN films were prepared by a non-solvent-induced phase separation process. The porous PAN films were irradiated with proton ions, and then carbonized in a tube furnace at various fluences under an inert atmosphere. It was found that pore structures were well maintained during proton irradiation and high-temperature carbonization processes. The Raman and XRD results showed that pseudo-graphitic structures were formed by the carbonization of proton irradiated porous PAN films. The electrical conductivity of the prepared PCFs increased with an increasing carbonization temperature due to the formation of more graphitic structures at a higher temperature.

  18. Arcing erosion of materials pre-irradiated with helium and heavy ions

    International Nuclear Information System (INIS)

    Nunogaki, M.; Yoshida, A.; Miyazaki, K.; Miyake, M.

    1985-01-01

    The surface erosion due to the action of high current arcs in a hydrogen plasma was investigated for aluminium, 316 stainless steel, titanium, molybdenum, tungsten and graphite. The effects of pre-irradiation of atomic deuterium ions, helium ions and self-ions related to the test materials on the erosion rate were investigated. The erosion was estimated by measuring the mass loss of arcing trace with use of a surface roughness gauge. The results show that the erosion rate decreased approximately with the melting temperature of materials with and without pre-irradiation. The erosion rates with pre-irradiation were increased, comparing to each erosion rate without any irradiation. The erosion rates were enhanced in the range of about 35% (tungsten) - 222% (aluminium). Some results of the erosion rate on the case of molybdenum films coated on graphite are described for discussion

  19. Arcing erosion of materials pre-irradiated with helium and heavy ions

    International Nuclear Information System (INIS)

    Nunogaki, M.; Yoshida, A.; Miyazaki, K.; Miyake, M.

    1984-01-01

    The surface erosion due to the action of high current arcs in a hydrogen plasma was investigated for aluminium, 316 stainless steel, titanium, molybdenum, tungsten and graphite. The effects of pre-irradiation of atomic deuterium ions, helium ions and self-ions related to the test materials on the erosion rate were investigated. The erosion was estimated by measuring the mass loss of arcing trace with use of a surface roughness gauge. The results show that the erosion rate decreased approximately with the melting temperature of materials with and without pre-irradiation. The erosion rates with pre-irradiation were increased, comparing to each erosion rate without any irradiation. The erosion rates were enhanced in the range of about 35 %(tungsten) - 222 %(aluminium). Some results of the erosion rate on the case of molybdenum films coated on graphite are described for discussion. (author)

  20. Fabrication of TREAT Fuel with Increased Graphite Loading

    Energy Technology Data Exchange (ETDEWEB)

    Luther, Erik Paul [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Leckie, Rafael M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dombrowski, David E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Papin, Pallas A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-02-05

    As part of the feasibility study exploring the replacement of the HEU fuel core of the TREAT reactor at Idaho National Laboratory with LEU fuel, this study demonstrates that it is possible to increase the graphite content of extruded fuel by reformulation. The extrusion process was use to fabricate the “upgrade” core1 for the TREAT reactor. The graphite content achieved is determined by calculation and has not been measured by any analytical method. In conjunction, a technique, Raman Spectroscopy, has been investigated for measuring the graphite content. This method shows some promise in differentiating between carbon and graphite; however, standards that would allow the technique to be calibrated to quantify the graphite concentration have yet to be fabricated. Continued research into Raman Spectroscopy is on going. As part of this study, cracking of graphite extrusions due to volatile evolution during heat treatment has been largely eliminated. Continued research to optimize this extrusion method is required.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-01

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

  2. Mode II interlaminar fracture of graphite/epoxy and graphite/PEEK

    Science.gov (United States)

    Carlsson, L. A.; Gillespie, J. W.; Trethewey, B. R.

    1986-01-01

    The end notched flexure (ENF) specimen is employed in an investigation of the interlaminar fracture toughness in Mode II (skew symmetric shear) loading of unidirectional graphite/epoxy and graphite/PEEK composites. Important experimental parameters such as the influence of precracking and the data reduction scheme for the Mode II toughness are discussed. Nonlinear load-deflection response is significant for the tough thermoplastic resin composite but is also present for the brittle thermoset composite. The observed nonlinearities, which are highly rate dependent, are attributed to a combination of slow stable crack growth preceding unstable crack growth and material inelastic behavior in the process zone around the crack tip.

  3. Food irradiation: An update

    International Nuclear Information System (INIS)

    Morrison, Rosanna M.

    1984-01-01

    Recent regulatory and commercial activity regarding food irradiation is highlighted. The effects of irradiation, used to kill insects and microorganisms which cause food spoilage, are discussed. Special attention is given to the current regulatory status of food irradiation in the USA; proposed FDA regulation regarding the use of irradiation; pending irradiation legislation in the US Congress; and industrial applications of irradiation

  4. 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

  5. Vermicular graphite cast iron current state of the art

    OpenAIRE

    Murthy, VSR; Seshan, S; Seshan, K

    1985-01-01

    Vermicular graphite cast iron is a new addition to the family of cast irons. Various methods for producing vermicular graphite cast iron are briefly discussed in this paper. The mechanical and physical properties of cast irons with vermicular graphite have been found to be intermediate between those of gray and ductile irons. Other properties such as casting characteristics, scaling resistance, damping capacity and machinability have been compared with those of gray and ductile irons. Probabl...

  6. Huge magnetoresistance effect of highly oriented pyrolytic graphite

    International Nuclear Information System (INIS)

    Du Youwei; Wang Zhiming; Ni Gang; Xing Dingyu; Xu Qingyu

    2004-01-01

    Graphite is a quasi-two-dimensional semimetal. However, for usual graphite the magnetoresistance is not so high due to its small crystal size and no preferred orientation. Huge positive magnetoresistance up to 85300% at 4.2 K and 4950% at 300 K under 8.15 T magnetic field was found in highly oriented pyrolytic graphite. The mechanism of huge positive magnetoresistance is not only due to ordinary magnetoresistance but also due to magnetic-field-driven semimetal-insulator transition

  7. Graphitic Carbon Nitride Supported Catalysts for Polymer Electrolyte Fuel Cells.

    OpenAIRE

    Mansor, N.; Jorge, A. B.; Corà, F.; Gibbs, C.; Jervis, R.; McMillan, P. F.; Wang, X.; Brett, D. J.

    2014-01-01

    Graphitic carbon nitrides are investigated for developing highly durable Pt electrocatalyst supports for polymer electrolyte fuel cells (PEFCs). Three different graphitic carbon nitride materials were synthesized with the aim to address the effect of crystallinity, porosity, and composition on the catalyst support properties: polymeric carbon nitride (gCNM), poly(triazine) imide carbon nitride (PTI/Li(+)Cl(-)), and boron-doped graphitic carbon nitride (B-gCNM). Following accelerated corrosion...

  8. Graphite structure and its relation to mechanical engineering design

    International Nuclear Information System (INIS)

    Brocklehurst, J.E.; Kelly, B.T.

    1980-01-01

    The inhomogeneous nature of polycrystalline graphite requires property measurements to be made over dimensions large enough to average the local variations in the structure. This is particularly true for mechanical integrity, and experimental data are presented which illustrate the importance of the real aggregate structure of graphite and the difficulties of interpreting strength data from different tests. The classical statistical treatments do not hold generally, and the problem of defining a failure criterion for graphite is discussed. It is suggested that the stress conditions in graphite components might be classified in terms of the dimensions and stress gradients related to the characteristic flaw size of the material as determined experimentally. (author)

  9. A study on wear behaviour of Al/6101/graphite composites

    Directory of Open Access Journals (Sweden)

    Pardeep Sharma

    2017-03-01

    Full Text Available The current research work scrutinizes aluminium alloy 6101-graphite composites for their mechanical and tribological behaviour in dry sliding environments. The orthodox liquid casting technique had been used for the manufacturing of composite materials and imperilled to T6 heat treatment. The content of reinforcement particles was taken as 0, 4, 8, 12 and 16 wt.% of graphite to ascertain it is prospective as self-lubricating reinforcement in sliding wear environments. Hardness, tensile strength and flexural strength of cast Al6101 metal matrix and manufactured composites were evaluated. Hardness, tensile strength and flexural strength decreases with increasing volume fraction of graphite reinforcement as compared to cast Al6101 metal matrix. Wear tests were performed on pin on disc apparatus to assess the tribological behaviour of composites and to determine the optimum volume fraction of graphite for its minimum wear rate. Wear rate reduces with increase in graphite volume fraction and minimum wear rate was attained at 4 wt.% graphite. The wear was found to decrease with increase in sliding distance. The average co-efficient of friction also reduces with graphite addition and its minimum value was found to be at 4 wt.% graphite. The worn surfaces of wear specimens were studied through scanning electron microscopy. The occurrence of 4 wt.% of graphite reinforcement in the composites can reveal loftier wear possessions as compared to cast Al6101 metal matrix.

  10. Dependence of strength on particle size in graphite

    International Nuclear Information System (INIS)

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

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

  11. 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)

  12. Thermal shock testing of ceramics with pulsed laser irradiation

    International Nuclear Information System (INIS)

    Benz, R.; Naoumidis, A.; Nickel, H.

    1986-04-01

    Arguments are presented showing that the resistance to thermal stressing (''thermal shock'') under pulsed thermal energy deposition by various kinds of beam irradiations is approximately proportional to Φ a √tp, where Φ a is the absorbed power density and tp is the pulse length, under conditions of diffusivity controlled spreading of heat. In practical beam irradiation testing, incident power density, Φ, is reported. To evaluate the usefulness of Φ√tp as an approximation to Φ a √tp, damage threshold values are reviewed for different kinds of beams (electron, proton, and laser) for a range of tp values 5x10 -6 to 2 s. Ruby laser beam irradiation tests were made on the following ceramics: AlN, BN, graphite, αSiC, β-SiC coated graphites, (α+β)Si 3 N 4 , CVD (chemical vapor deposition) TiC coated graphite, CVD TiC coated Mo, and CVD TiN coated IN 625. The identified failure mechanisms are: 1. plastic flow followed by tensile and bend fracturing, 2. chemical decomposition, 3. melting, and 4. loss by thermal spallation. In view of the theoretical approximations and the neglect of reflection losses there is reasonable accord between the damage threshold Φ√tp values from the laser, electron, and proton beam tests. (orig./IHOE)

  13. Food irradiation

    International Nuclear Information System (INIS)

    Beerens, H.; Saint-Lebe, L.

    1979-01-01

    Various aspects of food treatment by cobalt 60 or caesium 137 gamma radiation are reviewed. One of the main applications of irradiation on foodstuffs lies in its ability to kill micro-organisms, lethal doses being all the lower as the organism concerned is more complex. The effect on parasites is also spectacular. Doses of 200 to 300 krad are recommended to destroy all parasites with no survival period and no resistance phenomenon has ever been observed. The action of gamma radiation on macromolecules was also investigated, the bactericide treatment giving rise to side effects by transformation of food components. Three examples were studied: starch, nucleic acids and a whole food, the egg. The organoleptic aspect of irradiation was examined for different treated foods, then the physical transformations of unpasteurized, heat-pasteurized and radio-pasteurized eggs were compared. The report ends with a brief analysis of the toxicity and conditions of application of the treatment [fr

  14. AGC-2 Irradiation Data Qualification Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Laurence C. Hull

    2012-07-01

    The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The second Advanced Graphite Creep (AGC) experiment (AGC-2) began with Advanced Test Reactor (ATR) Cycle 149A on April 12, 2011, and ended with ATR Cycle 151B on May 5, 2012. The purpose of this report is to qualify AGC-2 irradiation monitoring data following INL Management and Control Procedure 2691, Data Qualification. Data that are Qualified meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Data that do not meet the requirements are Failed. Some data may not quite meet the requirements, but may still provide some useable information. These data are labeled as Trend. No Trend data were identified for the AGC-2 experiment. All thermocouples functioned throughout the AGC-2 experiment. There was one instance where spurious signals or instrument power interruption resulted in a recorded temperature value being well outside physical reality. This value was identified and labeled as Failed data. All other temperature data are Qualified. All helium and argon gas flow data are within expected ranges. Total gas flow was approximately 50 sccm through the capsule. Helium gas flow was briefly increased to 100 sccm during reactor shutdown. All gas flow data are Qualified. At the start of the experiment, moisture in the outflow gas line increased to 200 ppmv then declined to less than 10 ppmv over a period of 5 days. This increase in moisture coincides with the initial heating of the experiment and drying of the system. Moisture slightly exceeded 10 ppmv three other times during the experiment. While these moisture values exceed the 10 ppmv threshold value, the reported measurements are considered accurate and to reflect moisture conditions in the capsule. All moisture data are Qualified. Graphite creep specimens are subjected to one of three loads, 393 lbf

  15. The construction of the graphite calorimeter GR9 at LNE-LNHB (geometrical and technical consideration); Construction du calorimetre en graphite GR9 au LNE-LNHB (caracteristiques techniques et geometriques)

    Energy Technology Data Exchange (ETDEWEB)

    Ostrowsky, A.; Daures, J

    2008-07-01

    Calorimetry is the most direct dosimetric technique to reach absorbed dose. A calorimeter gives direct access to the energy imparted to matter by ionizing radiation per mass unit by measuring the heat quantity Q produced under irradiation in its sensitive element which is thermally insulated. Graphite was chosen as construction material because all the energy imparted to graphite by ionizing radiation is converted into heat. Thermistors are used for temperature measurements as well as for the electrical heating of the different bodies of the calorimeter. The construction of a calorimeter is the result of a compromise between dosimetric requirements and mechanical constraints. The difficulties encountered are examined and the solutions chosen are detailed. All technical data are gathered in this document. The aim is to provide a practical operative instruction and guidance document, which can help interested laboratories in designing such an instrument. The electrical and thermal tests have shown a good behaviour of the GR9 calorimeter.

  16. Surface areas of turbostratic graphitic carbons prepared from a resin using nickel particles, 20 nm, as graphitization catalyst

    Energy Technology Data Exchange (ETDEWEB)

    Oya, A.; Inoue, E.; Otani, S.; Marsh, H.

    1981-11-01

    Nickel particles were used to graphitize catalytically a non-graphitizing carbon to create a turbostratic graphitic material called the T/SUB/s-component. This method was examined by X-ray diffraction. Coals on heat treatment to temperatures >1270 K form T/SUB/s-component carbons. Therefore, considerations of the properties of the T/SUB/s-component carbon may have relevance to considerations of the operational performances of blast furnace coke. (22 refs.)

  17. Carbon/graphite composite material study. Appendix C: NASA studies on modification of carbon/graphite fibers and alternative materials

    Science.gov (United States)

    1981-01-01

    The feasibility of modifying resin matrix composites to reduce the potential of electrical shorting from fire released fiber was explored. The effort included modifications to or coatings for graphite fibers, alternative fibers, modifications to matrix materials, and hybrid composites. The objectives included reduction of the conductivity of the graphite fiber, char formation to reduce fiber release, glass formation to prevent fiber release, catalysis to assure fiber consumption in a fire, and replacement of the graphite fibers with nonconductive fibers of similar mechanical potential.

  18. Macroscopic Properties of Restacked, Redox-Liquid Exfoliated Graphite and Graphite Mimics Produced in Bulk Quantities

    Energy Technology Data Exchange (ETDEWEB)

    Srivastava, Vikram K [ORNL; Quinlan, Ronald [ORNL; Agapov, Alexander L [ORNL; Dunlap, John R [ORNL; Nelson, Kimberly M [ORNL; Duranty, Edward R [ORNL; Sokolov, Alexei P [ORNL; Bhat, Gajanan [ORNL; Mays, Jimmy [ORNL

    2014-01-01

    The excellent properties exhibited by monolayer graphene have spurred the development of exfoliation techniques using bulk graphite to produce large quantities of pristine monolayer sheets. Development of simple chemistry to exfoliate and intercalate graphite and graphite mimics in large quantities is required for numerous applications. To determine the macroscopic behavior of restacked, exfoliated bulk materials, a systematic approach is presented using a simple, redox-liquid sonication process along to obtain large quantities of 2D and 3D hexagonally layered graphite, molybdenum disulfi de, and boron nitride, which are subsequently characterized to observe chemical and structural changes. For MoS 2 sonicated with the antioxidant sodium bisulfi te, results from Raman spectroscopy, X-ray diffraction, and electron microscopy indicate the presence of distorted phases from different polymorphs, and apparent nanotube structures in the bulk, restacked powder. Furthermore, using thermograviemtric analysis, the antioxidant enhances the resistance to oxidative degradation of MoS 2 , upon thermal treatment up to 900 C. The addition of the ionic antioxidant decreased dispersion stability in non-polar solvent, suggesting decreased compatibility with non-polar systems. Using simple chemical methods, the ability to generate tailored multidimensional layered materials with unique macroscopic properties is critical for numerous applications, including electrical devices, reinforced polymer composites, lithium ion capacitors, and chemical sensing.

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

    Science.gov (United States)

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

    2016-03-01

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

  20. Graphite deposits in Siskiyou County, California

    Science.gov (United States)

    Rynearson, Garn A.

    1945-01-01

    The graphite deposits examined are in sec. 7, T. 47 N., R. 11 W., Siskiyou County, Calif., on the summit of the Siskiyou Mts. between Elk Meadow and the northeast end of "mill 6220" (see map of the Seias quadrangle). Four claims, designated as the Black Jack Nos. 1, 2, 3, and 4 claims, and a five-acre mill site have been located by W. H. Gassaway, W. B. Stewart, and E. R. Stewart. Development consists of four small cuts and several shallow trenches.

  1. Coating for gasifiable carbon-graphite fibers

    Science.gov (United States)

    Harper-Tervet, Jan (Inventor); Dowler, Warren L. (Inventor); Yen, Shiao-Ping S. (Inventor); Mueller, William A. (Inventor)

    1982-01-01

    A thin, uniform, firmly adherent coating of metal gasification catalyst is applied to a carbon-graphite fiber by first coating the fiber with a film-forming polymer containing functional moieties capable of reaction with the catalytic metal ions. Multivalent metal cations such as calcium cross-link the polymer such as a polyacrylic acid to insolubilize the film by forming catalytic metal macro-salt links between adjacent polymer chains. The coated fibers are used as reinforcement for resin composites and will gasify upon combustion without evolving conductive airborne fragments.

  2. Metallic Coatings for Graphite/Epoxy Composites

    Science.gov (United States)

    1980-08-01

    Perforated Foil-Coated Graphite/ Epoxy Panel .............. ........................... 4-6 4-4 Two-In. -Diameter 5052 Aluminum Alloy Repair Patch...The Phase I L i4-1 I1 evaluation indicated a need for a more corrosion-resistant aluminum alloy . The coating system selected was 0.0019-in. -thick 5052 ...application techniques on moisture resistance. The selected foil was 0.0025 in.-thick, 5052 aluminum alloy perforated with 0.010-in.-dlameter holes at a

  3. JACKETED FUEL ELEMENTS FOR GRAPHITE MODERATED REACTORS

    Science.gov (United States)

    Szilard, L.; Wigner, E.P.; Creutz, E.C.

    1959-05-12

    Fuel elements for a heterogeneous, fluid cooled, graphite moderated reactor are described. The fuel elements are comprised of a body of natural uranium hermetically sealed in a jacket of corrosion resistant material. The jacket, which may be aluminum or some other material which is non-fissionable and of a type having a low neutron capture cross-section, acts as a barrier between the fissioning isotope and the coolant or moderator or both. The jacket minimizes the tendency of the moderator and coolant to become radioactive and/or contaminated by fission fragments from the fissioning isotope.

  4. Irradiation test OF-2: high-temperature irradiation behavior of LASL-made fuel rods and LASL-made coated particles

    International Nuclear Information System (INIS)

    Wagner, P.; Reiswig, R.D.; Hollabaugh, C.M.; White, R.W.; O'Rourke, J.A.; Davidson, K.V.; Schell, D.H.

    1977-10-01

    Three LASL-made, substoichiometric ZrC-coated particles with inert kernels, and two high-density molded graphite fuel rods that contained LASL-made, ZrC-coated fissile particles were irradiated in the Oak Ridge Research Reactor test OF-2. The severest test conditions were 8.36 x 10 21 nvt (E greater than 0.18 MeV) at 1350 0 C. The graphite matrix showed no effect of the irradiation. There was no interaction between the matrix and any of the particle coats. The loose ZrC coated particles with inert kernels showed no irradiation effects. The graded ZrC-C coats on the fissile particles were cracked. It is postulated that the cracking is associated with the low LTI deposition rate and is not related to the ZrC

  5. Studies of defects on ion irradiated diamond

    Energy Technology Data Exchange (ETDEWEB)

    Lai, P.F.; Prawer, S.; Spargo, A.E.C.; Bursill, L.A. [Melbourne Univ., Parkville, VIC (Australia). School of Physics

    1996-12-31

    It is known that diamond is amorphized or graphitized when irradiated above a critical dose. Above this critical dose, D{sub c}, the resistance R is found to drop very rapidly due to the formation of graphite regions which overlap at D{sub c} to form a semi-continuous electrically conducting pathway through the sample. One particularly interesting method of studying this transformation is electron energy-loss spectroscopy (EELS). Using EELS, the different phases of carbon can be identified and distinguished from each other using the extended energy-loss fine structure (EXELFS) of the core-loss part of the spectrum. EELS is a sensitive method for determining the electronic structure of small areas of a sample. In this paper, transmission electron microscopy (TEM) and EELS measurements of the ion irradiated diamond were combined in an attempt to correlate the microstructural nature of the ion-beam induced damage to the changes in the electrical and other properties. 7 refs., 1 tab., 2 figs.

  6. Hydrogen retention properties of polycrystalline tungsten and helium irradiated tungsten

    International Nuclear Information System (INIS)

    Hino, T.; Koyama, K.; Yamauchi, Y.; Hirohata, Y.

    1998-01-01

    The hydrogen retention properties of a polycrystalline tungsten and tungsten irradiated by helium ions with an energy of 5 keV were examined by using an ECR ion irradiation apparatus and a technique of thermal desorption spectroscopy, TDS. The polycrystalline tungsten was irradiated at RT with energetic hydrogen ions, with a flux of 10 15 H cm -2 and an energy of 1.7 keV up to a fluence of 5 x 10 18 H cm -2 . Subsequently, the amount of retained hydrogen was measured by TDS. The heating temperature was increased from RT to 1000 C, and the heating rate was 50 C min -1 . Below 1000 C, two distinct hydrogen desorption peaks were observed at 200 C and 400 C. The retained amount of hydrogen was observed to be five times smaller than that of graphite, but the concentration in the implantation layer was comparable with that of graphite. Also, the polycrystalline tungsten was irradiated with 5 keV helium ions up to a fluence of 1.4 x 10 18 He cm -2 , and then re-irradiated with 1.7 keV hydrogen ions. The amount of retained hydrogen in this later experiment was close to the value in the case without prior helium ion irradiation. However, the amount of hydrogen which desorbed around the low temperature peak, 200 C, was largely enhanced. The desorption amount at 200 C saturated for the helium fluence of more than 5 x 10 17 He cm -2 . The present data shows that the trapping state of hydrogen is largely changed by the helium ion irradiation. Additionally, 5 keV helium ion irradiation was conducted on a sample pre-implanted with hydrogen ions to simulate a helium ion impact desorption of hydrogen retained in tungsten. The amount of the hydrogen was reduced as much as 50%. (orig.)

  7. Direct laser writing of nanorough cell microbarriers on anatase/Si and graphite/Si

    Energy Technology Data Exchange (ETDEWEB)

    Xiang, Y.C. [Departamento de Física Aplicada and Instituto Nicolás Cabrera, Universidad Autónoma de Madrid, Madrid (Spain); Department of Physics, Blackett Laboratory, Imperial College London, Prince Consort Road, London SW7 2BW (United Kingdom); Martínez-Martínez, R.M. [Departamento de Física Aplicada and Instituto Nicolás Cabrera, Universidad Autónoma de Madrid, Madrid (Spain); Torres-Costa, V. [Departamento de Física Aplicada and Instituto Nicolás Cabrera, Universidad Autónoma de Madrid, Madrid (Spain); Centro de Microanálisis de Materiales, Universidad Autónoma de Madrid, Madrid (Spain); Agulló-Rueda, F. [Instituto de Ciencia de Materiales de Madrid (ICMM-CSIC), 28049 Madrid (Spain); García-Ruiz, J.P. [Departamento de Biología Molecular, Universidad Autónoma de Madrid, 28049 Madrid (Spain); Manso-Silván, M., E-mail: miguel.manso@uam.es [Departamento de Física Aplicada and Instituto Nicolás Cabrera, Universidad Autónoma de Madrid, Madrid (Spain)

    2016-09-01

    The formation of hierarchical structures consisting of microstripe barriers decorated with nanorough ablated materials prepared by direct laser writing is described. Linear features of circa 25 μm width and 12 μm height are achieved on amorphous and crystalline titania and graphitic carbon films deposited on silicon. Ablated protrusions build up barriers decorated by nanoscale Si-film reconstructions, as indicated by EDX maps and micro-Raman spectroscopy. Wettability tests show a dramatic change in water contact angle, which leads to almost full wetting after irradiation, irrespective of the original film composition. Fluorescence microscopy images of human mesenchymal stem cells cultured on 1D and 2D structures demonstrate the short term biocompatibility of the ablated surfaces. It is shown that cells adhere, extend and polarize on feature edges, independently of the type of surface, thus suggesting that the created nanoroughness is at the origin of the antifouling behavior. In particular, irradiated anatase and graphite surfaces demonstrate an increased performance of crystalline films for the creation of cell guiding and trapping devices. The results suggest that such laser processing of films may serve as a time-and-cost-efficient method for the design of few-cells analytical surfaces. - Graphical abstract: Response of human mesenchymal stem cells to the microbarriers grown by direct laser writing on anatase/Si. Microbarriers show low colonization and high induction of cellular polarization on the feature edges. Display Omitted - Highlights: • Hierarchical microbarriers by direct IR laser writing on thin films. • Complex reconstruction of irradiated materials includes nanorough features. • Ablated areas become fully wettable. • Short term biocompatibility of the grown hierarchical structures confirmed. • Particular designs functional for cell guiding or trapping.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  9. Fabrication and irradiation performance of French HTGR fuel

    International Nuclear Information System (INIS)

    Moreau, C.; Ballagny, A.; Bujas, R.; Holder, J.; Ranc, R.; Soulhier, R.

    1977-01-01

    The fabrication of fuel for high-temperature reactors has been carried out in France either by using the techniques of the General Atomic Company or by adapting European techniques to the characteristics of the American fuel. Currently, the principal efforts are directed towards fuel containing enriched uranium and thorium. Prior to this, a large body of experience had been accumulated on the low enriched fuel of the type used in the Dragon reactor. Installations now in service permit fabrication, at realistic levels, of all the different parts comprising the fuel block. The kernels of fissile particles are made by loading of ion-exchange resin beads and conversion by thermal treatment. The fertile kernels are obtained by dry agglomeration of thorine powder. The coatings, of either BISO or TRISO design (pyrocarbon or pyrocarbon and silicon carbide), are deposited in a fluidized bed in a 127-mm-diameter industrial furnace. The fuel rods (or compacts) are obtained by injecting a pitch loaded with graphite powder into a bed of particles. Another technique, which is now under development, consists of hot-pressing the particles which have been mixed with resin-impregnated graphite powder and then polymerizing them. In both cases, a final thermal treatment is carried out at 1800 0 C. Research and development in the machining of the prismatic graphite blocks has been performed by the Atelier ''Graphite'' at the Marcoule Research Centre. A large programme of irradiations has permitted testing of the individual fuel components followed by tests on complete fuel assemblies. Particles, fuel rods and graphite have been tested in capsules placed in experimental French reactors. Fuel elements have been proof-tested either in the irradiation loops of the Osiris and Pegasus reactors or in the Dragon reactor. The results of these tests have shown that the French fuel can withstand the operating conditions in a power reactor (temperature, fast fluence, burnup) without exceeding the

  10. Measurement of the specific heat capacity of graphite

    International Nuclear Information System (INIS)

    Picard, S.; Burns, D.T.; Roger, P.

    2006-01-01

    With the objective of implementing graphite calorimetry at the BIPM to measure absorbed dose, an experimental assembly has recently been constructed to measure the specific heat capacity of graphite. A status description of the apparatus and results from the first measurements are given. The outcome is discussed and the experimental uncertainty is reviewed. (authors)

  11. 'Indicator'carbonaceous phyllite/graphitic schist in the Archean ...

    Indian Academy of Sciences (India)

    Carbonaceous rocks in the form of graphitic schist and carbonaceous phyllite are the major host rocks of the gold mineralization in Kundarkocha gold deposit of the Precambrian Singhbhum orogenic belt in eastern India. The detection of organic carbon, essentially in the carbonaceous phyllite and graphitized schist within ...

  12. Carboxylic acid terminated, solution exfoliated graphite by organic ...

    Indian Academy of Sciences (India)

    Graphite nanosheets are considered as a promising material for a range of applications from flexible electronics to functional nanodevices such as biosensors, intelligent coatings and drug delivery. Chemical functionalizationof graphite nanosheets with organic/inorganic materials offers an alternative approach to control ...

  13. Evaluation of ac conductivity behaviour of graphite filled

    Indian Academy of Sciences (India)

    Composites of epoxy resin having different amounts of graphite particles have been prepared by solution casting method. Temperature dependence of dielectric constant, tan and a.c. conductivity was measured in the frequency range, 1–20 kHz, temperature range, 40–180°C for 0.99, 1.96 and 2.91 wt% graphite filled ...

  14. Measurements of anomalous neutron transport in bulk graphite

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    Indian Academy of Sciences (India)

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

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

  16. Industry technology assessment of graphite-polymide composite materials. [conferences

    Science.gov (United States)

    1975-01-01

    An assessment of the current state of the art and the future prospects for graphite polyimide composite material technology is presented. Presentations and discussions given at a minisymposium of major issues on the present and future use, availability, processing, manufacturing, and testing of graphite polyimide composite materials are summarized.

  17. Linear expansion of products out of thermal splitting graphite

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  18. A TEM Study on the Microstructure of Fine Flaky Graphite

    DEFF Research Database (Denmark)

    Moumeni, Elham; Tiedje, Niels Skat; Horsewell, Andy

    In this investigation the microstructure of the graphite flakes in titanium alloyed cast iron is studied using electron microscopy techniques. Based on this information, growth models for the platelets in the fine graphite flakes in cast iron are considered. Detailed crystallographic analysis...

  19. AGC-3 Experiment Irradiation Monitoring Data Qualification Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Hull, Laurence C. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Technology Development Office

    2014-08-01

    The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear-grade graphite. The third experiment, Advanced Graphite Creep 3 (AGC-3), began with Advanced Test Reactor (ATR) Cycle 152B on November 27, 2012, and ended with ATR Cycle 155B on April 23, 2014. This report documents qualification of AGC-3 experiment irradiation monitoring data for use by the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Program for research and development activities required to design and license the first VHTR nuclear plant. Qualified data meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements. Trend data may not meet the requirements, but may still provide some useable information. The report documents qualification of AGC-3 experiment irradiation monitoring data following MCP-2691. This report also documents whether AGC-3 experiment irradiation monitoring data meet the requirements for data collection as specified in technical and functional requirements documents and quality assurance (QA) plans. Data handling is described showing how data are passed from the data collection experiment to the Nuclear Data Management and Analysis System (NDMAS) team. The data structure is described, including data batches, components, attributes, and response variables. The description of the approach to data qualification includes the steps taken to qualify the data and the specific tests used to verify that the data meet requirements. Finally, the current status of the data received by NDMAS from the AGC-3 experiment is presented with summarized information on test results and resolutions. This report addresses all of the irradiation monitoring data collected during the AGC-3 experiment.

  20. Graphitic Carbon Nitride Supported Catalysts for Polymer Electrolyte Fuel Cells

    Science.gov (United States)

    2014-01-01

    Graphitic carbon nitrides are investigated for developing highly durable Pt electrocatalyst supports for polymer electrolyte fuel cells (PEFCs). Three different graphitic carbon nitride materials were synthesized with the aim to address the effect of crystallinity, porosity, and composition on the catalyst support properties: polymeric carbon nitride (gCNM), poly(triazine) imide carbon nitride (PTI/Li+Cl–), and boron-doped graphitic carbon nitride (B-gCNM). Following accelerated corrosion testing, all graphitic carbon nitride materials are found to be more electrochemically stable compared to conventional carbon black (Vulcan XC-72R) with B-gCNM support showing the best stability. For the supported catalysts, Pt/PTI-Li+Cl– catalyst exhibits better durability with only 19% electrochemical surface area (ECSA) loss versus 36% for Pt/Vulcan after 2000 scans. Superior methanol oxidation activity is observed for all graphitic carbon nitride supported Pt catalysts on the basis of the catalyst ECSA. PMID:24748912

  1. Structures and Performance of Graphene/Polyimide Composite Graphite Fibers

    Directory of Open Access Journals (Sweden)

    LI Na

    2017-09-01

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

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

    International Nuclear Information System (INIS)

    Berry, S.M.

    1993-09-01

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

  3. Cluster Deposition and Implantation on/in Graphite

    DEFF Research Database (Denmark)

    Popok, Vladimir

    2013-01-01

    Cluster ion beam technique is a versatile tool which can be used for controllable formation of nanosize objects on the surface, modification and processing of surfaces and shallow layers on an atomic scale. In this chapter an overview of research on cluster interaction with graphite is presented....... One of the emphases is put on pinning of metal clusters on graphite with a possibility of following selective etching of graphene layers. The other topic of concern is related to the phenomenon of cluster stopping and the development of scaling law for cluster implantation in graphite. Graphite...... is chosen for surface experiments because it is a good model material; it has an atomically smooth surface that makes it easy to resolve very small deposited clusters or damaged areas. Layered structure of graphite with strong covalent bonds in the graphene sheets and very week van der Waals interactions...

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  5. Graphitic Carbon Nitride Supported Catalysts for Polymer Electrolyte Fuel Cells.

    Science.gov (United States)

    Mansor, Noramalina; Jorge, A Belen; Corà, Furio; Gibbs, Christopher; Jervis, Rhodri; McMillan, Paul F; Wang, Xiaochen; Brett, Daniel J L

    2014-04-03

    Graphitic carbon nitrides are investigated for developing highly durable Pt electrocatalyst supports for polymer electrolyte fuel cells (PEFCs). Three different graphitic carbon nitride materials were synthesized with the aim to address the effect of crystallinity, porosity, and composition on the catalyst support properties: polymeric carbon nitride (gCNM), poly(triazine) imide carbon nitride (PTI/Li + Cl - ), and boron-doped graphitic carbon nitride (B-gCNM). Following accelerated corrosion testing, all graphitic carbon nitride materials are found to be more electrochemically stable compared to conventional carbon black (Vulcan XC-72R) with B-gCNM support showing the best stability. For the supported catalysts, Pt/PTI-Li + Cl - catalyst exhibits better durability with only 19% electrochemical surface area (ECSA) loss versus 36% for Pt/Vulcan after 2000 scans. Superior methanol oxidation activity is observed for all graphitic carbon nitride supported Pt catalysts on the basis of the catalyst ECSA.

  6. INCREASING OF MECHANICAL PROPERTIES OF CAST GRAPHITIZED STEEL

    Directory of Open Access Journals (Sweden)

    I. V. Akimov

    2015-06-01

    Full Text Available Purpose. Due to the presence of many essential properties (low prime cost, high manufacturability and damping ability, heat conductivity, fluidity and others,the graphitized cast irons are widely applied for parts operating under the conditions of static and cyclic loads, wear at dry friction and aggressive media at elevated temperatures. At the same time because of significant content of the graphite phase in the structure, the cast irons possess relatively low mechanical properties. Thereby the attention is drawn to graphitized steels, the peculiarity of which consists in the presence of graphite inclusions in the structure, which appoints specific cast iron's properties to these materials. But unlike cast irons, the graphite content in them is in 2…3 times lower and it affects positively on mechanical and service properties indices. This work deals with the optimization of the composition of cast graphitized steel, which has high mechanical properties indices. Methodology. The experimental design technique was used in this work. Alloys prototypes were smelted and their mechanical properties were investigated on the basis of these techniques. Findings were exposed to regression processing, and the dependences of the alloys components influence on its properties were obtained. By means of graphical optimization the optimal composition of steel with high mechanical properties indices has been determined. Findings.Ithasbeenestablishedthat carbon and silicon have the most significant influence on the strength and cyclic endurance of graphitized steels. The chemical composition of the cast graphitized steel with high static and cyclic strength indices was suggested in this work. Originality. With the use of mathematical experimental design techniques the dependences describing the influence of carbon, silicon and copper on the static and cyclic strength indices were obtained. They allow optimizing compositions of graphitized steels. Practical

  7. Rate-dependent mode I interlaminar crack growth mechanisms in graphite/epoxy and graphite/PEEK

    Science.gov (United States)

    Gillespie, J. W., Jr.; Carlsson, L. A.; Smiley, A. J.

    1987-01-01

    In this paper the mode I fracture behavior of graphite/epoxy and graphite/PEEK composites is examined over four decades of crosshead rates (0.25-250 mm/min). Straight-sided double-cantilever-beam specimens consisting of unidirectional laminates were tested at room temperature. For graphite/epoxy the load-deflection response was linear to fracture, and stable slow crack growth initiating at the highest load level was observed for all rates tested. In contrast, mode I crack growth in the graphite/PEEK material was often unstable and showed stick-slip behavior. Subcritical crack growth occurring prior to the onset of fracture was observed at intermediate displacement rates. A mechanism for the fracture behavior of the graphite/PEEK material (based on viscoelastic, plastic, and microcrack coalescence in the process zone) is proposed and related to the observed rate-dependent phenomena.

  8. Irradiation growth in zirconium alloys: a review

    International Nuclear Information System (INIS)

    Fidleris, V.

    1980-09-01

    The change in shape during irradiation without external stress, irradiation growth, was first discovered in uranium and later in graphite, zirconium and other core materials which exhibit anisotropic physical properties. The direction of maximum growth of metals invariably corresponds with the direction of minimum thermal expansion. In polycrystalline zirconium alloys growth is positive in the direction of maximum deformation during fabrication and in other directions it can be either positive or negative depending on the preferred orientation of grains (crystallographic texture). Growth increases gradually with temperature between 300 K and 620 K and rapidly with fluence up to about 1 x 10 25 n.m. -2 (Eμ1 MeV). At higher fluences the growth appears to saturate in annealed materials and reach a steady rate approximately proportional to dislocation density in cold-worked materials. Above 600 K both annealed and cold-worked materials have similar steady growth rates. Irradiation growth is caused by the segregation to different sinks of the vacancies and interstitials generated by irradiation, but the dominant types of sinks for each type of point defect and the mode of transport of the point defects to sinks cannot therefore be predicted theoretically. For the purpose of designing reactor core components empirical equations have been derived that can satisfactorily predict the steady state growth behaviour from texture and microstructure. (auth)

  9. STIGMA STIG STEGT STAGT STABA, Stress Analysis of Dragon HTR Graphite Structure

    International Nuclear Information System (INIS)

    Kinkead, A.N.

    2002-01-01

    1 - Nature of the physical problem solved: Stress analysis of graphite structures for the DRAGON high temperature reactor is performed by this family of computer codes. Two-dimensional plane strain irradiation dose dependent core problems have been solved. 2 - Method of solution: STAGT, which is the oldest in this series of programmes, can handle multiply connected regions but is confined to plane strain in x-y geometry. Variations in temperature loading during irradiation is accounted for (Wigner strain component.) STIG, is a version of STAGT where an anisotropic elasticity matrix has been introduced to handle transversely isotropic materials. An additional feature of 'STIG' is the introduction of a boundary restraint condition of practical importance to prismatic gas cooled reactor core construction. This is defined as rotational plane strain in which free distortion of the prism arising from overall gradient of temperature and/or fast neutron damage flux coincident with any single direction may be assumed to occur if variation of thermal expansion coefficient with irradiation is included. 'STIGMA' is intended for evaluation of stress and displacement in composite axisymmetrical bodies subject to variable loadings in the axial and radial directions. The code has been prepared to take account of transverse isotropy in material characteristics for up to four separate bonded interface zones within a single composite material problem. Although specifically designed for the analysis of graphite structural components in the fast neutron irradiation environment of a reactor core, it is equally applicable to initial state design of prestressed concrete pressure vessels and other problems involving rotational symmetry. 'STABA'-stress,temperature and bowing analysis. The aim of this quasi 3-D computer code is to apply the principle of rotational plane strain over the full length of a prismatic core component, taking into account spatial variations in fast neutron and

  10. Study of hot corrosion of flakes of non purified graphite and of purified graphite

    International Nuclear Information System (INIS)

    Boule, Michel

    1967-01-01

    The author reports the study of hot corrosion of the Ticonderoga graphite. He reports the study of the defects of graphite flakes (structure defects due to impurities), the dosing of these impurities, and then their removal by purification. Flakes have then been oxidised by means of a specially designed apparatus. Based on photographs taken by optical and electronic microscopy, the author compares the oxidation features obtained in dry air and in humid air, between purified and non purified flakes. He also reports the study of the evolution of oxidation with respect to the initial rate of impurities, and the study of the evolution of oxidation features in humid air during oxidation. All these comparisons are made while taking the oxidation rate into account [fr

  11. Investigation of irradiated X-ray crystal stability for ITER diagnostics system

    Energy Technology Data Exchange (ETDEWEB)

    Kosenkov, V.M. (V.I. Lenin Research Institute of Atomic Reactors, 433510 Dimitrovgrad (Russian Federation)); Neverov, V.A. (V.I. Lenin Research Institute of Atomic Reactors, 433510 Dimitrovgrad (Russian Federation)); Revyakin, Y.L. (V.I. Lenin Research Institute of Atomic Reactors, 433510 Dimitrovgrad (Russian Federation)); Orlinsky, D.V. (Russian Scientific Center, Kurchatov Institute, 123182 Moscow (Russian Federation))

    1994-09-01

    The possible neutron irradiation effect on gamma-spectrometer operation intended for diagnostics of ITER plasma was analysed. Nine types of crystals Ge (111), Si (220), SiO[sub 2] (1010), SiO[sub 2] (1120), SiO[sub 2] (0003), pyrolytic graphite (0002), pentaeritrit (002), potassium biphtalate (001), rubidium biphtalate (001) used in monochromator were irradiated in the BOR-60 reactor up to fluence 2.2x10[sup 19] n/cm[sup -2] at temperature 65 C. Change of reflecting ability, interplane distance, width and form for diffraction reflections were studied. Ge, Si and graphite crystals were not changed after irradiation. In SiO[sub 2] the damage depends on reflection plane. The organic crystals were delaninated and cracked. ((orig.))

  12. 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.

  13. 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.

  14. Design Procedure of Graphite Components by ASME HTR Codes

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  15. Air oxidation behavior of carbon and graphite materials for HTGR

    International Nuclear Information System (INIS)

    Kawakami, Haruo

    1986-01-01

    Most components in the core of high temperature gas-cooled reactors are made of carbon and graphite which are efficient neutron moderators, and have high strength at high temperature. The demerit of these materials in HTGR use is that these are readily oxidized by the impurity oxidants in helium coolant in the normal operating condition, and by air in the case of an air ingress accident. In order to examine the candidate materials for the experimental very high temperature gas-cooled reactor in Japan, the air oxidation experiment on some carbon and graphite was carried out. The materials tested were isotropic fine grain graphite (1G-11, 1G-110), anisotropic molded graphite (PGX, TS-1621), and anisotropic molded carbon (ASR-ORB, ASR-IRB, P3JHA-B). The uniform oxidation in the temperature range from 430 to 650 deg C and the non-uniform oxidation in the temperature range from 700 to 1000 deg C were tested. The oxidation of graphite by air was enhanced by the impurities in the graphite such as Co, Ni and V. The reaction rate of PGX graphite was nearly proportional to oxygen partial pressure. Below 650 deg C, the ratio of reaction products CO/CO 2 increased as temperature rose, but above 800 deg C, CO was oxidized to CO 2 . (Kako, I.)

  16. On the relation between strength properties of carbon materials and total porosity following neutron irradiation

    International Nuclear Information System (INIS)

    Virgil'ev, Yu.S.; Makarchenko, V.G.

    1981-01-01

    Correlation of compression strength and flexural strength and dynamic elastic modulus with the general porosity of carbon materials irradiated with neutrons is considered. The syntactic foam and fine-grain graphite on the base of non-calcinated oil coke with pitch binding have been used as samples for investigations. The irradiation is carried out by neutron fluence of 1.3x10 21 neutr/cm 2 . It is shown that the above correlation as well as in non-irradiated materials is approximated satisfactorily by the exponential equation. Besides the character of the interaction of filler grains between themselves and with the binding, has not changed [ru

  17. Automatic classification of graphite in cast iron.

    Science.gov (United States)

    Gomes, Otávio da F M; Paciornik, Sidnei

    2005-08-01

    A method for automatic classification of the shape of graphite particles in cast iron is proposed. In a typical supervised classification procedure, the standard charts from the ISO-945 standard are used as a training and validation population. Several shape and size parameters are described and used as discriminants. A new parameter, the average internal angle, is proposed and is shown to be relevant for accurate classification. The ideal parameter sets are determined, leading to validation success rates above 90%. The classifier is then applied to real cast iron samples and provides results that are consistent with visual examination. The method provides classification results per particle, different from the traditional per field chart comparison methods. The full procedure can run automatically without user interference.

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

    Energy Technology Data Exchange (ETDEWEB)

    Stansfield, O.M.

    1990-07-25

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

  19. Chemical analysis of high purity graphite

    International Nuclear Information System (INIS)

    1993-03-01

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

  20. Absorption spectra of AA-stacked graphite

    International Nuclear Information System (INIS)

    Chiu, C W; Lee, S H; Chen, S C; Lin, M F; Shyu, F L

    2010-01-01

    AA-stacked graphite shows strong anisotropy in geometric structures and velocity matrix elements. However, the absorption spectra are isotropic for the polarization vector on the graphene plane. The spectra exhibit one prominent plateau at middle energy and one shoulder structure at lower energy. These structures directly reflect the unique geometric and band structures and provide sufficient information for experimental fitting of the intralayer and interlayer atomic interactions. On the other hand, monolayer graphene shows a sharp absorption peak but no shoulder structure; AA-stacked bilayer graphene has two absorption peaks at middle energy and abruptly vanishes at lower energy. Furthermore, the isotropic features are expected to exist in other graphene-related systems. The calculated results and the predicted atomic interactions could be verified by optical measurements.