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

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.

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

Diffusion of cesium and iodine in compressed IG-110 graphite compacts

International Nuclear Information System (INIS)

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

2016-01-01

Nuclear graphite grade IG-110 is currently used in the High Temperature Engineering Test Reactor (HTTR) in Japan for certain permanent and replaceable core components, and is a material of interest in general. Therefore, transport parameters for fission products in this material are needed. Measurement of diffusion through pressed compacts of IG-110 graphite is experimentally attractive because they are easy to prepare with homogeneous distributions of fission product surrogates. In this work, we measured diffusion coefficients for Cs and I in pressed compacts made from IG-110 powder in the 1079–1290 K temperature range, and compared them to those obtained in as-received IG-110. - Highlights: • A method for analysis of fission product diffusion in graphite by ICP-MS was applied to pressed IG-110 graphite compacts containing cesium and iodine. • Diffusion coefficients for cesium and iodine were obtained. • The measurement design simulates HTGR conditions of high temperature and flowing helium.

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

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

International Nuclear Information System (INIS)

Yoda, S.; Eto, M.

1983-01-01

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

Statistics of fracture in two grades of isotropic graphite

International Nuclear Information System (INIS)

Kennedy, C.R.; Montgomery, S.C.

1990-01-01

Properties in large billets of graphite used in critical applications demand a high level of quality assurance. Therefore, it must be determined if sample test results represent the properties of the entire billet. Flexure tests were performed on specimens from 27 populations in each grade with respect to billet, position, and orientation to establish the confidence levels for estimation of overall variance and mean strength in the billet. Comparisons of tensile and brittle ring to flexure strengths were made. Homogeneity of variance was found to be a tenable hypothesis; however, estimates of the billet mean strength were not as confidently predicted by the samples. (orig.)

Statistics of fracture in two grades of isotropic graphite

Science.gov (United States)

Kennedy, C. R.; Montgomery, S. C.

1990-04-01

Properties in large billets of graphite used in critical applications demand a high level of quality assurance. Therefore, it must be determined if sample test results represent the properties of the entire billet. Flexure tests were performed on specimens from 27 populations in each grade with respect to billet, position, and orientation to establish the confidence levels for estimation of overall variance and mean strength in the billet. Comparisons of tensile and brittle ring to flexure strengths were made. Homogeneity of variance was found to be a tenable hypothesis; however, estimates of the billet mean strength were not as confidently predicted by the samples.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-03-15

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

Nondestructive evaluation of nuclear-grade graphite

Science.gov (United States)

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

2012-05-01

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

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

Effects of Oxidation on Oxidation-Resistant Graphite

Energy Technology Data Exchange (ETDEWEB)

Windes, William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Rebecca [Idaho National Lab. (INL), Idaho Falls, ID (United States); Carroll, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-05-01

The Advanced Reactor Technology (ART) Graphite Research and Development Program is investigating doped nuclear graphite grades that exhibit oxidation resistance through the formation of protective oxides on the surface of the graphite material. In the unlikely event of an oxygen ingress accident, graphite components within the VHTR core region are anticipated to oxidize so long as the oxygen continues to enter the hot core region and the core temperatures remain above 400°C. For the most serious air-ingress accident which persists over several hours or days the continued oxidation can result in significant structural damage to the core. Reducing the oxidation rate of the graphite core material during any air-ingress accident would mitigate the structural effects and keep the core intact. Previous air oxidation testing of nuclear-grade graphite doped with varying levels of boron-carbide (B4C) at a nominal 739°C was conducted for a limited number of doped specimens demonstrating a dramatic reduction in oxidation rate for the boronated graphite grade. This report summarizes the conclusions from this small scoping study by determining the effects of oxidation on the mechanical strength resulting from oxidation of boronated and unboronated graphite to a 10% mass loss level. While the B4C additive did reduce mechanical strength loss during oxidation, adding B4C dopants to a level of 3.5% or more reduced the as-fabricated compressive strength nearly 50%. This effectively minimized any benefits realized from the protective film formed on the boronated grades. Future work to infuse different graphite grades with silicon- and boron-doped material as a post-machining conditioning step for nuclear components is discussed as a potential solution for these challenges in this report.

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.

Graphite selection for the FMIT test cell

International Nuclear Information System (INIS)

Morgan, W.C.

1982-06-01

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

Final report on graphite irradiation test OG-2

International Nuclear Information System (INIS)

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

1975-01-01

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

Thermal decomposition of cesium-ethylene-ternary graphite intercalation compounds

International Nuclear Information System (INIS)

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

2010-01-01

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

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

ICT media design for higher grade of elementary school mathematics learning using CS6 program

Science.gov (United States)

Zainil, M.; Prahmana, R. C. I.; Helsa, Y.; Hendri, S.

2017-12-01

Technological innovation contributes to the emerging of new possibilities to change the learning process. The development of technology could bring the higher quality of education through the integration of technology in the learning. The purpose of this research is to create an interactive multimedia using CS6 program for mathematics learning in higher grade of elementary school. It was a development research using ADDIE model which consists of analysis, design, and evaluation stages. It has successfully developed interactive multimedia in a form of learning CD used in the material of plane figures and solid figures. The prototype has been validated and then tested for the 4th grade of elementary schools. Two schools were involved and the students taught by utilizing the prototype, and then, in the end of learning, they are examined to determine the learning result. There were 72% of the students passed the examination as they classified at good and excellent categories. Finally, the use of CS6 program is promising to help the students learning plane and solid figure in mathematics learning.

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

Free vibration of fully functionally graded carbon nanotube reinforced graphite/epoxy laminates

Science.gov (United States)

Kuo, Shih-Yao

2018-03-01

This study provides the first-known vibration analysis of fully functionally graded carbon nanotube reinforced hybrid composite (FFG-CNTRHC) laminates. CNTs are non-uniformly distributed to reinforce the graphite/epoxy laminates. Some CNT distribution functions in the plane and thickness directions are proposed to more efficiently increase the stiffening effect. The rule of mixtures is modified by considering the non-homogeneous material properties of FFG-CNTRHC laminates. The formulation of the location dependent stiffness matrix and mass matrix is derived. The effects of CNT volume fraction and distribution on the natural frequencies of FFG-CNTRHC laminates are discussed. The results reveal that the FFG layout may significantly increase the natural frequencies of FFG-CNTRHC laminate.

Microstructural characterization and pore structure analysis of nuclear graphite

International Nuclear Information System (INIS)

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

2011-01-01

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

Floatability study of graphite ore from southeast Sulawesi (Indonesia)

Science.gov (United States)

Florena, Fenfen Fenda; Syarifuddin, Fahmi; Hanam, Eko Sulistio; Trisko, Nici; Kustiyanto, Eko; Enilisiana, Rianto, Anton; Arinton, Ghenadi

2016-02-01

Graphite ore obtained from Kolaka Regency, South East Sulawesi, Indonesia have been succesfully investigated for beneficiation by froth flotation technique. Preliminary study have been done to determine the minerals types, fixed carbon content and liberation size of the graphite. Graphite is naturally floatable due to its hydrophobic property. Some suitable reagents are usually added to increase effectiveness of recovery. In this article, enrichment of graphite by froth flotation was studied by investigating the effect of reagents concentrations, rotation speed and particle size on the carbon grade and recovery of the concentrate. The carbon grade increased from 3.00% to 60.00% at the optimum flotation conditions.

Corrosion-induced microstructural changes in a US core graphite

International Nuclear Information System (INIS)

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

1981-01-01

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

Effect of various dopant elements on primary graphite growth

International Nuclear Information System (INIS)

Valle, N; Theuwissen, K; Lacaze, J; Sertucha, J

2012-01-01

Five spheroidal graphite cast irons were investigated, a usual ferritic grade and four pearlitic alloys containing Cu and doped with Sb, Sn and Ti. These alloys were remelted in a graphite crucible, leading to volatilization of the magnesium added for spheroidization and to carbon saturation of the liquid. The alloys were then cooled down and maintained at a temperature above the eutectic temperature. During this step, primary graphite could develop showing various features depending on the doping elements added. The largest effects were that of Ti which greatly reduces graphite nucleation and growth, and that of Sb which leads to rounded agglomerates instead of lamellar graphite. The samples have been investigated with secondary ion mass spectrometry to enlighten distribution of elements in primary graphite. SIMS analysis showed almost even distribution of elements, including Mg and Al (from the inoculant) in the ferritic grade, while uneven distribution was evident in all doped alloys. Investigations are going on to clarify if the uneven distribution is associated with structural defects in the graphite precipitates.

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

Energy Technology Data Exchange (ETDEWEB)

Mark C. Carroll; David T. Rohrbaugh

2013-08-01

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

Graphite for high-temperature reactors

International Nuclear Information System (INIS)

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

1976-01-01

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

Feasibility of monitoring the strength of HTGR core support graphite. Part II

International Nuclear Information System (INIS)

Morgan, W.C.; Becker, F.L.

1979-08-01

The results reported establish the technical feasibility of a method for monitoring the strength of HTGR core support structures in situ. Correlations have been established between the velocity of an ultrasonic pulse and the compressive strength of four different grades of graphite. For some grades of graphite, one or more of the correlations are practically independent of oxidation profile in samples having cylindrical geometry (as in the core support posts). For other grades of graphite, and for other sample geometries, the oxidation-depth profile must be known in order to reliably predict the effect of oxidation on compressive strength

Graphite content and isotopic fractionation between calcite-graphite pairs in metasediments from the Mgama Hills, Southern Kenya

International Nuclear Information System (INIS)

Arneth, J.D.; Schidlowski, M.; Sarbas, B.; Goerg, U.; Amstutz, G.C.

1985-01-01

Amphibolite-grade metasediments from the Mgama Hills region, Kenya, contain conspicuous quantities of graphite, most probably derived from organic progenitor materials,. The highest graphite contents are found in schists whereas calcite marbles intercalated in the sequence contain relatively low amounts. The graphitic constituents are consistently enriched in 13 C relative to common sedimentary organic material, with the highest isotopic ratios in graphite from the marbles. Carbon isotope fractionations between calcite and graphite mostly vary between 3.3 and 7.1 per mille, which comes close to both empirically recorded and thermodynamically calculated fractionations in the temperature range of the upper amphibolite facies. However, larger values occasionally encountered in the marbles suggest that complete isotopic equilibrium is not always attained in amphibolite-facies metamorphism. (author)

Oxidation of PCEA nuclear graphite by low water concentrations in helium

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

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

International Nuclear Information System (INIS)

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

2012-01-01

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

The Role of Cesium Cation in Controlling Interphasial Chemistry on Graphite Anode in Propylene Carbonate-Rich Electrolytes

Energy Technology Data Exchange (ETDEWEB)

Xiang, Hongfa; Mei, Donghai; Yan, Pengfei; Bhattacharya, Priyanka; Burton, Sarah D.; Cresce, Arthur V.; Cao, Ruiguo; Engelhard, Mark H.; Bowden, Mark E.; Zhu, Zihua; Polzin, Bryant; Wang, Chong M.; Xu, Kang; Zhang, Jiguang; Xu, Wu

2015-09-10

Propylene carbonate (PC) is seldom used in lithium-ion batteries (LIBs) due to its sustained co-intercalation into graphene structure and the eventual graphite exfoliation, despite potential advantages it brings, such as wider liquid range and lower cost. Here we discover that cesium cation (Cs+), originally used to suppress dendrite growth of Li metal anode, directs the formation of solid electrolyte interphase (SEI) on graphitic anode in PC-rich electrolytes through preferential solvation. Effective suppression of PC-decomposition and graphite-exfoliation was achieved when the ratio of ethylene carbonate (EC)/PC in electrolytes was so adjusted that the reductive decomposition of Cs+-(EC)m (1≤m≤2) complex precedes that of Li+-(PC)n (3≤n≤5). The interphase directed by Cs+ is stable, ultrathin and compact, leading to significant improvements in LIB performances. In a broader context, the accurate tailoring of SEI chemical composition by introducing a new solvation center represents a fundamental breakthrough in manipulating interfacial reactions processes that once were elusive.

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

Why are women underrepresented in Computer Science? Gender differences in stereotypes, self-efficacy, values, and interests and predictors of future CS course-taking and grades

Science.gov (United States)

Beyer, Sylvia

2014-07-01

This study addresses why women are underrepresented in Computer Science (CS). Data from 1319 American first-year college students (872 female and 447 male) indicate that gender differences in computer self-efficacy, stereotypes, interests, values, interpersonal orientation, and personality exist. If students had had a positive experience in their first CS course, they had a stronger intention to take another CS course. A subset of 128 students (68 females and 60 males) took a CS course up to one year later. Students who were interested in CS, had high computer self-efficacy, were low in family orientation, low in conscientiousness, and low in openness to experiences were more likely to take CS courses. Furthermore, individuals who were highly conscientious and low in relational-interdependent self-construal earned the highest CS grades. Efforts to improve women's representation in CS should bear these results in mind.

Utilization of Cs137 to generate a radiation barrier for weapons grade plutonium immobilized in borosilicate glass canisters. Revision 1

International Nuclear Information System (INIS)

Jardine, L.J.; Armantrout, G.A.; Collins, L.F.

1995-01-01

One of the ways recommended by a recent National Academy of Sciences study to dispose of excess weapons-grade plutonium is to encapsulate the plutonium in a glass in combination with high-level radioactive wastes (HLW) to generate an intense radiation dose rate field. The objective is to render the plutonium as difficult to access as the plutonium contained in existing US commercial spent light-water reactor (LWR) fuel until it can be disposed of in a permanent geological repository. A radiation dose rate from a sealed canister of 1,000 rem/h (10 Sv/h) at 1 meter for at least 30 years after fabrication was assumed in this paper to be a radiation dose comparable to spent LWR fuel. This can be achieved by encapsulating the plutonium in a borosilicate glass with an adequate amount of a single fission product in the HLWS, namely radioactive Cs 137 . One hundred thousand curies of Cs 137 will generate a dose rate of 1,000 rem/h (10 Sv/h) at 1 meter for at least 30 years when imbedded into canisters of the size proposed for the Savannah River Site's vitrified high-level wastes. The United States has a current inventory of 54 MCi of CS 137 that has been separated from defense HLWs and is in sealed capsules. This single curie inventory is sufficient to spike 50 metric tons of excess weapons-grade plutonium if plutonium can be loaded at 5.5 wt% in glass, or 540 canisters. Additional CS 137 inventories exist in the United States' HLWs from past reprocessing operations, should additional curies be required. Using only one fission product, CS 137 , rather than the multiple chemical elements and compounds in HLWs to generate a high radiation dose rate from a glass canister greatly simplifies the processing engineering retirement for encapsulating plutonium in a borosilicate glass

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

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.

Simulating Neutron Radiation Damage of Graphite by In-situ Electron Irradiation

International Nuclear Information System (INIS)

Mironov, Brindusa E; Freeman, H M; Brydson, R M D; Westwood, A V K; Scott, A J

2014-01-01

Radiation damage in nuclear grade graphite has been investigated using transmission electron microscopy (TEM) and electron energy loss spectroscopy (EELS). Changes in the structure on the atomic scale and chemical bonding, and the relationship between each were of particular interest. TEM was used to study damage in nuclear grade graphite on the atomic scale following 1.92×10 8 electrons nm −2 of electron beam exposure. During these experiments EELS spectra were also collected periodically to record changes in chemical bonding and structural disorder, by analysing the changes of the carbon K-edge. Image analysis software from the 'PyroMaN' research group provides further information, based on (002) fringe analysis. The software was applied to the micrographs of electron irradiated virgin 'Pile Grade A' (PGA) graphite to quantify the extent of damage from electron beam exposure

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

International Nuclear Information System (INIS)

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

1975-01-01

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

Special graphites; Graphites speciaux

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [French] Ameliorer les proprietes du graphite nucleaire pour empilements et ouvrir de nouveaux domaines d'application au graphite constituent une part importante de l'effort entrepris en commun par le Commissariat a l'Energie Atomique (CEA) et la compagnie PECHINEY. Des procedes nouveaux de fabrication de carbones et graphites speciaux ont ete mis au point: graphite forge, pyrocarbone, graphite de haute densite, agglomeration de poudres de graphite par craquage de gaz naturel, graphites impermeables. Les proprietes physiques de ces produits ainsi que leur reaction avec differents gaz oxydants sont decrites. Les premiers resultats d'irradiation sont aussi donnes. (auteurs)

Analysis of DTI-Derived Tensor Metrics in Differential Diagnosis between Low-grade and High-grade Gliomas.

Science.gov (United States)

Jiang, Liang; Xiao, Chao-Yong; Xu, Quan; Sun, Jun; Chen, Huiyou; Chen, Yu-Chen; Yin, Xindao

2017-01-01

Purpose: It is critical and difficult to accurately discriminate between high- and low-grade gliomas preoperatively. This study aimed to ascertain the role of several scalar measures in distinguishing high-grade from low-grade gliomas, especially the axial diffusivity (AD), radial diffusivity (RD), planar tensor (Cp), spherical tensor (Cs), and linear tensor (Cl) derived from diffusion tensor imaging (DTI). Materials and Methods: Fifty-three patients with pathologically confirmed brain gliomas (21 low-grade and 32 high-grade) were included. Contrast-enhanced T1-weighted images and DTI were performed in all patients. The AD, RD, Cp, Cs, and Cl values in the tumor zone, peritumoral edema zone, white matter (WM) adjacent to edema and contralateral normal-appearing white matter (NAWM) were calculated. The DTI parameters and tumor grades were statistically analyzed, and receiver operating characteristic (ROC) curve analysis was also performed. Results: The DTI metrics in the affected hemisphere showed significant differences from those in the NAWM, except for the AD values in the tumor zone and the RD values in WM adjacent to edema in the low-grade groups, as well as the Cp values in WM adjacent to edema in the high-grade groups. AD in the tumor zone as well as Cs and Cl in WM adjacent to edema revealed significant differences between the low- and high-grade gliomas. The areas under the curve (Az) of all three metrics were greater than 0.5 in distinguishing low-grade from high-grade gliomas by ROC curve analysis, and the best DTI metric was Cs in WM adjacent to edema (Az: 0.692). Conclusion: AD in the tumor zone as well as Cs and Cl in WM adjacent to edema will provide additional information to better classify gliomas and can be used as non-invasive reliable biomarkers in glioma grading.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-01

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

Carbon Papers and Aerogels Based on Graphene Layers and Chitosan: Direct Preparation from High Surface Area Graphite.

Science.gov (United States)

Barbera, Vincenzina; Guerra, Silvia; Brambilla, Luigi; Maggio, Mario; Serafini, Andrea; Conzatti, Lucia; Vitale, Alessandra; Galimberti, Maurizio

2017-12-11

In this work, carbon papers and aerogels based on graphene layers and chitosan were prepared. They were obtained by mixing chitosan (CS) and a high surface area nanosized graphite (HSAG) in water in the presence of acetic acid. HSAG/CS water dispersions were stable for months. High resolution transmission electron microscopy revealed the presence of few graphene layers in water suspensions. Casting or lyophilization of such suspensions led to the preparation of carbon paper and aerogel, respectively. In X-ray spectra of both aerogels and carbon paper, peaks due to regular stacks of graphene layers were not detected: graphene with unaltered sp 2 structure was obtained directly from graphite without the use of any chemical reaction. The composites were demonstrated to be electrically conductive thanks to the graphene. Chitosan thus makes it possible to obtain monolithic carbon aerogels and flexible and free-standing graphene papers directly from a nanosized graphite by avoiding oxidation to graphite oxide and successive reduction. Strong interaction between polycationic chitosan and the aromatic substrate appears to be at the origin of the stability of HSAG/CS adducts. Cation-π interaction is hypothesized, also on the basis of X-ray photoelectron spectroscopy findings. This work paves the way for the easy large-scale preparation of carbon papers through a method that has a low environmental impact and is based on a biosourced polymer, graphene, and water.

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

Graphite development for gas-cooled reactors in the USA

International Nuclear Information System (INIS)

Burchell, T.D.

1991-01-01

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

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

International Nuclear Information System (INIS)

2014-01-01

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.

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.

The influence of roughness on tribological properties of nuclear grade graphite

International Nuclear Information System (INIS)

Luo Xiaowei; Yu Suyuan; Sheng Xuanyu; He Shuyan

2006-01-01

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

Brazing graphite to graphite

International Nuclear Information System (INIS)

Peterson, G.R.

1976-01-01

Graphite is joined to graphite by employing both fine molybdenum powder as the brazing material and an annealing step that together produce a virtually metal-free joint exhibiting properties similar to those found in the parent graphite. Molybdenum powder is placed between the faying surfaces of two graphite parts and melted to form molybdenum carbide. The joint area is thereafter subjected to an annealing operation which diffuses the carbide away from the joint and into the graphite parts. Graphite dissolved by the dispersed molybdenum carbide precipitates into the joint area, replacing the molybdenum carbide to provide a joint of graphite

Recent developments in graphite. [Use in HTGR and aerospace

Energy Technology Data Exchange (ETDEWEB)

Cunningham, J.E.

1983-01-01

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

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)

Characteristics of first loaded IG-110 graphite in HTTR core

International Nuclear Information System (INIS)

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

2006-10-01

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

Surface area-burnoff correlation for the steam--graphite reaction

International Nuclear Information System (INIS)

Stark, W.A. Jr.; Malinauskas, A.P.

1977-01-01

The oxidation of core graphite by steam of air represents a problem area of significant concern in safety analyses for the high temperature gas cooled reactor (HTGR). Core and core-support graphite integrity and strength deteriorate with oxidation of the graphite, and oxidation furthermore could affect the rate of fission product release under upset conditions. Consequently, modeling of core response during steam or air ingress conditions requires an expression for the rate of graphite interaction with those impurities. The steam--graphite reaction in particular is a complex interaction of mass transport within the graphite with chemi-sorption and reaction on accessible surfaces; experimental results from graphite to graphite are highly variable, and the description of the reaction is not yet completely consistent. A simple etch pit model relating surface area to burnoff has been proposed and shown to provide reasonable correlation with experimental data obtained from steam oxidation studies of nuclear grade H-327 graphite. Unaccounted differences between theory and experiment arise at burnoffs exceeding 3 to 5 percent. The model, while not complete nor comprehensive, is consistent with experimental observations of graphite oxidation by O 2 (air), CO 2 , or H 2 O, and could have some utility in safety analysis

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Graphite core design in UK reactors

International Nuclear Information System (INIS)

Davies, M.W.

1996-01-01

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

The effect of oxidizing atmosphere on strength loss in HTGR graphites

International Nuclear Information System (INIS)

Heiser, J.H.; Finfrock, C.C.; Lees, B.S.

1983-01-01

This paper reports on studies involving various reactor grade graphites and the possible mechanisms leading to strength loss differences. Compressive and tensile specimens of six reactor grade graphites were oxidized. The compressive or tensile strengths were then determined using a Timus-Olsen Universal testing machine following ASTM standard test specifications. Two possible mechanisms are proposed to explain the differences in strength loss given the same mass loss but different oxidants. One mechanism has the impurity iron located primarily in the filler particles and the second mechanism arranges the iron either uniformly throughout the binder or inhomogeneously dispersed in large pockets in the binder

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.

Low cost sic coated erosion resistant graphite

International Nuclear Information System (INIS)

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

2007-01-01

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

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

A graphite foam reinforced by graphite particles

Energy Technology Data Exchange (ETDEWEB)

Zhu, J.J.; Wang, X.Y.; Guo, L.F.; Wang, Y.M.; Wang, Y.P.; Yu, M.F.; Lau, K.T.T. [DongHua University, Shanghai (China). College of Material Science and Engineering

2007-11-15

Graphite foam was obtained after carbonization and graphitization of a pitch foam formed by the pyrolysis of coal tar based mesophase pitch mixed with graphite particles in a high pressure and temperature chamber. The graphite foam possessed high mechanical strength and exceptional thermal conductivity after adding the graphite particles. Experimental results showed that the thermal conductivity of modified graphite foam reached 110W/m K, and its compressive strength increased from 3.7 MPa to 12.5 MPa with the addition of 5 wt% graphite particles. Through the microscopic observation, it was also found that fewer micro-cracks were formed in the cell wall of the modified foam as compared with pure graphite foam. The graphitization degree of modified foam reached 84.9% and the ligament of graphite foam exhibited high alignment after carbonization at 1200{sup o}C for 3 h and graphitization at 3000{sup o}C for 10 min.

Effect of friction on oxidative graphite intercalation and high-quality graphene formation.

Science.gov (United States)

Seiler, Steffen; Halbig, Christian E; Grote, Fabian; Rietsch, Philipp; Börrnert, Felix; Kaiser, Ute; Meyer, Bernd; Eigler, Siegfried

2018-02-26

Oxidative wet-chemical delamination of graphene from graphite is expected to become a scalable production method. However, the formation process of the intermediate stage-1 graphite sulfate by sulfuric acid intercalation and its subsequent oxidation are poorly understood and lattice defect formation must be avoided. Here, we demonstrate film formation of micrometer-sized graphene flakes with lattice defects down to 0.02% and visualize the carbon lattice by transmission electron microscopy at atomic resolution. Interestingly, we find that only well-ordered, highly crystalline graphite delaminates into oxo-functionalized graphene, whereas other graphite grades do not form a proper stage-1 intercalate and revert back to graphite upon hydrolysis. Ab initio molecular dynamics simulations show that ideal stacking and electronic oxidation of the graphite layers significantly reduce the friction of the moving sulfuric acid molecules, thereby facilitating intercalation. Furthermore, the evaluation of the stability of oxo-species in graphite sulfate supports an oxidation mechanism that obviates intercalation of the oxidant.

Independent yields of Rb and Cs isotopes from thermal-neutron induced fission of 235U

International Nuclear Information System (INIS)

Balestrini, S.J.; Decker, R.; Wollnik, H.; Wuensch, K.D.; Jung, G.; Koglin, E.; Siegert, G.

1979-01-01

The relative yields of Rb and Cs isotopes from thermal-neutron fission of 235 U have been redetermined using the mass separator OSTIS, on-line at a neutron guide of the High-Flux Beam Reactor at the Institut Laue-Langevin, Grenoble, France. The separator ion source was a hot oven containing 235 U in a graphite matrix. The neutron beam was pulsed. Alkali fission products diffused out of the graphite and were ionized, thus producing a stepwise increase in the analyzed ion beam proportional to the independent fission yield. The ion beam and the fissions in the source were monitored simultaneously. The diffusion of Rb and Cs from the source was exponential in time with half-lives ranging from 2.8 to 18 sec, depending upon the element and source temperature. The independent fission yields of Rb and Cs are normalized by equating their element yields to each other and to a value computed from the charge distributions observed with the recoil separator LOHENGRIN and well established mass yields. Fractional independent yields are deduced from the independent fission yields, and these compare very well with the EOZ model described by Wahl

Interpretation of bend strength increase of graphite by the couple-stress theory

International Nuclear Information System (INIS)

Tang, P.Y.

1981-05-01

This paper presents a continued evaluation of the applicability of the couple-stress constitutive theory to graphite. The evaluation is performed by examining four-point bend and uniaxial tensile data of various sized cylindrical and square specimens for three grades of graphites. These data are superficially inconsistent and, usually, at variance with the predictions of classical theories. Nevertheless, this evaluation finds that they can be consistently interpreted by the couple-stress theory. This is compatible with results of an initial evaluation that considered one size of cylindrical specimen for H-451 graphite

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)

Acceptance test for graphite components and construction status of HTTR

International Nuclear Information System (INIS)

Iyoku, T.; Ishihara, M.; Maruyama, S.; Shiozawa, S.; Tsuji, N.; Miki, T.

1996-01-01

In March, 1991, the Japan Atomic Energy Research Institute (JAERI) started to constructed the High Temperature engineering Test Reactor(HTTR) which is a 30-MW(thermal) helium gas-cooled reactor with a core composed of prismatic graphite blocks piled on the core support graphite structures. Two types of graphite materials are used in the HTTR. One is the garde IG-110, isotropic fine grain graphite, another is the grade PGX, medium-to-fine grained molded graphite. These materials were selected on the basis of the appropriate properties required by the HTTR reactor design. Industry-wide standards for an acceptance test of graphite materials used as main components of a nuclear reactor had not been established. The acceptance standard for graphite components of the HTTR, therefore, was drafted by JAERI and reviewed by specialists outside JAERI. The acceptance standard consists of the material testing, non-destructive examination such as the ultrasonic and eddy current testings, dimensional and visual inspections and assembly test. Ultrasonic and eddy current testings are applied to graphite logs to detect an internal flaw and to graphite components to detect a surface flaw, respectively. The assembly test is performed at the works, prior to their installation in the reactor pressure vessel, to examine fabricating precision of each component and alignment of piled-up structures. The graphite components of the HTTR had been tested on the basis of the acceptance standard. It was confirmed that the graphite manufacturing process was well controlled and high quality graphite components were provided to the HTTR. All graphite components except for the fuel graphite blocks are to be installed in the reactor pressure vessel of the HTTR in September 1995. The paper describes the construction status of the HTTR focusing on the graphite components. The acceptance test results are also presented in this paper. (author). Figs

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

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.

Independent yields of Rb and Cs isotopes from thermal-neutron induced fission of /sup 235/U

Energy Technology Data Exchange (ETDEWEB)

Balestrini, S.J.; Decker, R.; Wollnik, H.; Wuensch, K.D.; Jung, G.; Koglin, E.; Siegert, G.

1979-12-01

The relative yields of Rb and Cs isotopes from thermal-neutron fission of /sup 235/U have been redetermined using the mass separator OSTIS, on-line at a neutron guide of the High-Flux Beam Reactor at the Institut Laue-Langevin, Grenoble, France. The separator ion source was a hot oven containing /sup 235/U in a graphite matrix. The neutron beam was pulsed. Alkali fission products diffused out of the graphite and were ionized, thus producing a stepwise increase in the analyzed ion beam proportional to the independent fission yield. The ion beam and the fissions in the source were monitored simultaneously. The diffusion of Rb and Cs from the source was exponential in time with half-lives ranging from 2.8 to 18 sec, depending upon the element and source temperature. The independent fission yields of Rb and Cs are normalized by equating their element yields to each other and to a value computed from the charge distributions observed with the recoil separator LOHENGRIN and well established mass yields. Fractional independent yields are deduced from the independent fission yields, and these compare very well with the EOZ model described by Wahl.

Conditioning for definitive storage of radioactive graphite bricks from reactor decommissioning

International Nuclear Information System (INIS)

Costes, J.R.; Koch, C.; Tassigny, C. de; Vidal, H.; Raymond, A.

1990-01-01

The decommissioning of gas-graphite reactors in the EC (e.g. French UNGGs, British Magnox reactors and AGRs, and reactors in Spain and in Italy) will produce large amounts of graphite bricks. This graphite cannot be accepted without particular conditioning by the existing shallow land disposal sites. The aim of the study is to examine the behaviour of graphite waste and to develop a conditioning technique which makes this waste acceptable for shallow land disposal sites. 18 kg of graphite core samples with an outside diameter of 74 mm were removed from the G2 gas-cooled reactor at Marcoule. Their radioactivity is highly dependent on the position of the graphite bricks inside the reactor. Measured results indicate an activity range of 100-400 MBq/kg with 90% Tritium, 5% 14 C, 3% 60 Co, 1.5% 63 Ni. Repeated porosity analyses showed that open porosity ranging from 0 to 100 μm exceeded 23 vol% in the graphite. Water penetration kinetics were investigated in unimpregnated graphite and resulted in impregnation by water of 50-90% of the open porosity. Preliminary lixiviation tests on the crude samples showed quick lixidegree of Cs (several per cent) and of 60 Co, and 133 Ba at a lesser degree. The proposed conditioning technique does not involve a simple coating but true impregnation by a tar-epoxy mixture. The bricks recovered intact from the core by robot services will be placed one by one inside a cylindrical metallic container. But this container may corrode and the bricks may become fragmented in the future, the normally porous graphite will be unaffected by leaching since it is proved that all pores larger than 0.1 μm will be filled with the tar-epoxy mixture. This is a true long-term waste packaging concept. The very simple technology required for industrial implementation is discussed

76 FR 14910 - Small Diameter Graphite Electrodes From the People's Republic of China: Initiation of Anti...

Science.gov (United States)

2011-03-18

... correct grade of petroleum coke mix, and have been baked, formed, carbonized, impregnated, and graphitized... essential characteristics of a graphite electrode; (c) UKCG's own proprietary description of the U.K... are performed, which do not impart any essential performance characteristics to the finished product...

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Effect of Different Concrete Grade on Radiation Linear Attenuation Coefficient (μ)

International Nuclear Information System (INIS)

Noor Azreen Masenwat; Mohammad Shahrizan Samsu; Mohamad Pauzi Ismail; Suhairy Sani; Ismail Mustapha; Nasharuddin Isa; Mohamad Haniza Mahmud

2014-01-01

In calculating the quantity of absorption of radiation and its relationship with the thickness of a material, linear attenuation coefficient (μ) of the material is one of the parameters to be taken into account. For normal concrete, the (μ) varies depending on the type of radiation used, 0.105 cm -1 for Co-60 and 0.123 cm -1 for Cs-137. Value (μ) is used in the calculation of the radiation absorption for concrete material does not take into account factors such concrete grades. In this research, concrete with different grades (Grade 15, Grade 20, Grade 25, Grade 30, Grade 35, Grade 40) are designed and manufactured with reference to the mixing method described in British Standard. Then, the linear attenuation (μ) for each grade are measured using the radiation from the source Co-60 and Cs-137 sources. This paper describes and discusses the impact of differences in concrete grade of linear attenuation (μ) for Co-60 source/ source Cs-137 and its relationship with the compressive strength. (author)

Thermal fatigue of refractory metal/graphite composites for fusion applications

International Nuclear Information System (INIS)

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

1995-01-01

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

Graphite to Inconel brazing using active filler metal

International Nuclear Information System (INIS)

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

1989-01-01

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

Surface heterogeneity and ionization of Cs promoter in carbon-based ruthenium catalyst for ammonia synthesis

International Nuclear Information System (INIS)

Kotarba, Andrzej; Dmytrzyk, Jaromir; Rarog-Pilecka, Wioletta; Kowalczyk, Zbigniew

2003-01-01

Second-generation ammonia synthesis cesium-doped ruthenium catalyst supported on turbostratic carbon was investigated by the species resolved thermal alkali desorption method (SR-TAD). Energetic barriers for cesium ions (2.86 eV), ground state (1.96 eV) and electronically excited atoms (5.76 eV) desorbing from the Cs-Ru/C catalyst were determined. In the case of ruthenium-free Cs/C system, cesium desorbs as ground state atoms only, with an energy barrier of 2.87 eV. The work functions determined by the thermionic emission of electrons from Cs/C and Cs-Ru/C were of the same value (2.9 eV). It was concluded that ruthenium induces heterogeneous distribution of cesium on the catalyst surface. The promoter stability is reduced on low work function areas and its surface ionization on high work function areas opens the ionic desorption channel. The Cs desorption from the catalyst is discussed in terms of the literature data for the cesium/graphite system

Fission Product Sorptivity in Graphite

Energy Technology Data Exchange (ETDEWEB)

Tompson, Jr., Robert V. [Univ. of Missouri, Columbia, MO (United States); Loyalka, Sudarshan [Univ. of Missouri, Columbia, MO (United States); Ghosh, Tushar [Univ. of Missouri, Columbia, MO (United States); Viswanath, Dabir [Univ. of Missouri, Columbia, MO (United States); Walton, Kyle [Univ. of Missouri, Columbia, MO (United States); Haffner, Robert [Univ. of Missouri, Columbia, MO (United States)

2015-04-01

graduate student meant that data acquisition with the packed bed systems ended up competing for the graduate student’s available time with the electrodynamic balance redesign and assembly portions of the project. This competition for available time was eventually mitigated to some extent by the later recruitment of an undergraduate student to help with data collection using the packed bed system. It was only the recruitment of the second student that allowed the single particle balance design and construction efforts to proceed as far as they did during the project period. It should be added that some significant time was also spent by the graduate student cataloging previous work involving graphite. This eventually resulted in a review paper being submitted and accepted (“Adsorption of Iodine on Graphite in High Temperature Gas-Cooled Reactor Systems: A Review,” Kyle L. Walton, Tushar K. Ghosh, Dabir S. Viswanath, Sudarshan K. Loyalka, Robert V. Tompson). Our specific revised objectives in this project were as follows: Experimentally obtain isotherms of Iodine for reactor grade IG-110 samples of graphite particles over a range of temperatures and pressures using an EDB and a temperature controlled EDB; Experimentally obtain isotherms of Iodine for reactor grade IG-110 samples of graphite particles over a range of temperatures and pressures using a packed column bed apparatus; Explore the effect that charge has on the adsorption isotherms of iodine by varying the charges on and the voltages used to suspend the microscopic particles in the EDB; and To interpret these results in terms of the existing models (Langmuir, BET, Freundlich, and others) which we will modify as necessary to include charge related effects.

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

Studies of the role of molten materials in interactions with UO2 and graphite

International Nuclear Information System (INIS)

Fink, J.K.; Heiberger, J.J.; Leibowitz, L.

1979-01-01

Graphite, which is being considered as a lower reactor shield in gas-cooled fast reactors, would be contacted by core debris during a core disruptive accident. Information on the interaction of graphite, UO 2 , and stainless steel is needed in assessing the safety of the GCFR. In an ongoing study of the interaction of graphite, UO 2 , and stainless steel, the effects of the steel components have been investigated by electron microprobe scans, x-ray diffraction, and reaction-rate measurements. Experiments to study the role of the reaction product, FeUC 2 , in the interaction suggested that FeUC 2 promotes the interaction by acting as a carrier to bring graphite to the reaction site. Additional experiments using pyrolytic graphite show that while the reaction rate is decreased at 2400 K, at higher temperatures the rate is similar to that using other grades of graphite

Graphite nanoreinforcements in polymer nanocomposites

Science.gov (United States)

Fukushima, Hiroyuki

, or carbon blacks. The exfoliated graphite flakes reached the percolation threshold at 1.93 wt% (1.13 vol%) in an epoxy system and the resistivity of the composite showed 39 ohm•cm with 7 wt% of exfoliated graphite, which is comparable to the high-grade carbon black based systems. The vapor grown carbon fiber based composites showed higher resistivity at the same filler contents while the conventional carbon fiber composites showed much higher resistivity and percolation threshold. Stress distribution analysis by Finite Element Method revealed the stress concentration condition of composite systems is affected by factors such as shape of the reinforcements, aspect ratio, and geological arrangements. Based on these results, an optimal morphology design of nanocomposite system was proposed. Market research revealed that there is a realistic possibility for applying the new process and material in commercial products and a venture business plan was proposed based on this new technology. The venture plan won "The Most Innovative Design" award at the 2002 Michigan Collegiate Entrepreneur's Conference.

Evaluation of high temperature brazes for graphite first wall protection elements

International Nuclear Information System (INIS)

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

1995-01-01

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

Thermal fatigue of refractory metal / graphite composites for fusion applications

International Nuclear Information System (INIS)

Smid, I.; Nickel, H.

1989-01-01

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

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.

Experience of on-site disposal of production uranium-graphite nuclear reactor.

Science.gov (United States)

Pavliuk, Alexander O; Kotlyarevskiy, Sergey G; Bespala, Evgeny V; Zakharova, Elena V; Ermolaev, Vyacheslav M; Volkova, Anna G

2018-04-01

The paper reported the experience gained in the course of decommissioning EI-2 Production Uranium-Graphite Nuclear Reactor. EI-2 was a production Uranium-Graphite Nuclear Reactor located on the Production and Demonstration Center for Uranium-Graphite Reactors JSC (PDC UGR JSC) site of Seversk City, Tomsk Region, Russia. EI-2 commenced its operation in 1958, and was shut down on December 28, 1990, having operated for the period of 33 years all together. The extra pure grade graphite for the moderator, water for the coolant, and uranium metal for the fuel were used in the reactor. During the operation nitrogen gas was passed through the graphite stack of the reactor. In the process of decommissioning the PDC UGR JSC site the cavities in the reactor space were filled with clay-based materials. A specific composite barrier material based on clays and minerals of Siberian Region was developed for the purpose. Numerical modeling demonstrated the developed clay composite would make efficient geological barriers preventing release of radionuclides into the environment. Copyright © 2018 Elsevier Ltd. All rights reserved.

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Structural disorder of graphite and implications for graphite thermometry

Science.gov (United States)

Kirilova, Martina; Toy, Virginia; Rooney, Jeremy S.; Giorgetti, Carolina; Gordon, Keith C.; Collettini, Cristiano; Takeshita, Toru

2018-02-01

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

Asymptomatic Intracorneal Graphite Deposits following Graphite Pencil Injury

OpenAIRE

Philip, Swetha Sara; John, Deepa; John, Sheeja Susan

2012-01-01

Reports of graphite pencil lead injuries to the eye are rare. Although graphite is considered to remain inert in the eye, it has been known to cause severe inflammation and damage to ocular structures. We report a case of a 12-year-old girl with intracorneal graphite foreign bodies following a graphite pencil injury.

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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.

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

International Nuclear Information System (INIS)

Hong, Jin Ki; Chi, Se Hwan

2006-01-01

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

Evaluation of high temperature brazes for graphite first wall protection elements

International Nuclear Information System (INIS)

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

1989-01-01

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

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

International Nuclear Information System (INIS)

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

1988-04-01

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

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%

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

Science.gov (United States)

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

2013-06-25

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-04-15

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

Theoretical study of the Cs isotope exchange reaction of CsI + Cs' → Cs + ICs' (Contract research)

International Nuclear Information System (INIS)

Kobayashi, Takanori; Hashimoto, Masashi; Yokoyama, Keiichi

2015-12-01

To discuss the exchange reaction of Cs isotope by CsI + Cs' → Cs + ICs', the structure and chemical properties of Cs 2 I intermediate and potential energy surface of the entrance reaction are calculated using M06/def2-TZVPPD density functional calculation. The calculation shows that the reaction to the intermediate has no barrier and the two Cs-I bonds of Cs 2 I are chemically equivalent. These results suggest that the rate of the Cs exchange reaction of CsI + Cs' → Cs + ICs' is as high as the collision rate. (author)

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)

High-resolution optical microscopy of carbon and graphite

International Nuclear Information System (INIS)

Cook, W.H.; Allen, M.D.; Leslie, B.C.; Gray, R.J.

1975-01-01

The ceramographic preparation of carbonaceous materials varying in crystalline quality, amorphous carbon to well crystallized graphite, is described. In a two-step process, using alumina and diamond polishing compounds, one can prepare more samples, obtain a substantial saving in man hours, avoid rounding material around pores, and obtain flatter surfaces than were obtainable with earlier, conventional methods. Improved resolution of microstructural details is achieved without impregnation with epoxy resins or other materials to support the porous structures. Use of rotatable, half-wave retardation (sensitive tint) enhances the microstructural definition in both color and black and white. These innovations were extensively used as part of the examination of nuclear grades of graphite before and after exposure to fast neutrons at temperatures from 650 to 1100 0 C; typical examples are discussed. (auth)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-01

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

Determination of gold and cobalt dopants in advanced materials based on tin oxide by slurry sampling high-resolution continuum source graphite furnace atomic absorption spectrometry

Science.gov (United States)

Filatova, Daria G.; Eskina, Vasilina V.; Baranovskaya, Vasilisa B.; Vladimirova, Svetlana A.; Gaskov, Alexander M.; Rumyantseva, Marina N.; Karpov, Yuri A.

2018-02-01

A novel approach is developed for the determination of Co and Au dopants in advanced materials based on tin oxide using high-resolution continuum source graphite furnace atomic absorption spectrometry (HR CS GFAAS) with direct slurry sampling. Sodium carboxylmethylcellulose (Na-CMC) is an effective stabilizer for diluted suspensions. Use Na-CMC allows to transfer the analytes into graphite furnace completely and reproducibly. The relative standard deviation obtained by HR CS GFAAS was not higher than 4%. Accuracy was proven by means inductively coupled plasma mass spectrometry (ICP-MS) in solutions after decomposition as a comparative technique. To determine Au and Co in the volume of SnO2, the acid decomposition conditions (HCl, HF) of the samples were suggested by means of an autoclave in a microwave oven.

Graphite curtain vacuum outgassing and heat transfer. Final report

International Nuclear Information System (INIS)

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

1976-12-01

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

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.

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

International Nuclear Information System (INIS)

Mohanty, Subhasish; Majumdar, Saurindranath; Srinivasan, Makuteswara

2013-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-15

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Process for purifying graphite

International Nuclear Information System (INIS)

Clausius, R.A.

1985-01-01

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

Purification and preparation of graphite oxide from natural graphite

Energy Technology Data Exchange (ETDEWEB)

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

2016-03-11

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-08-14

Superhydrophilic graphite surfaces and water-dispersible graphite colloids are obtained by electrochemical exfoliation with hydrophobic graphite electrodes. Such counterintuitive characteristics are caused by partial oxidation and investigated by examining both graphite electrodes and exfoliated particles after electrolysis. The extent of surface oxidation can be explored through contact angle measurement, scanning electron microscope, electrical sheet resistance, x-ray photoelectron spectroscopy, zeta-potential analyzer, thermogravimetric analysis, UV-visible, and Raman spectroscopy. The degree of wettability of the graphite anode can be altered by the electrolytic current and time. The water contact angle declines generally with increasing the electrolytic current or time. After a sufficient time, the graphite anode becomes superhydrophilic and its hydrophobicity can be recovered by peeling with adhesive tape. This consequence reveals that the anodic graphite is oxidized by oxygen bubbles but the oxidation just occurs at the outer layers of the graphite sheet. Moreover, the characteristics of oxidation revealed by UV peak shift, peak ratio between D and G bands, and negative zeta-potential indicate the presence of graphite oxide on the outer shell of the exfoliated colloids. However, thermogravimetric analysis for the extent of decomposition of oxygen functional groups verifies that the amount of oxygen groups is significantly less than that of graphite oxide prepared via Hummer method. The structure of this partially oxidized graphite may consist of a graphite core covered with an oxidized shell. The properties of the exfoliated colloids are also influenced by pH of the electrolytic solution. As pH is increased, the extent of oxidation descends and the thickness of oxidized shell decreases. Those results reveal that the degree of oxidation of exfoliated nanoparticles can be manipulated simply by controlling pH.

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

International Nuclear Information System (INIS)

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

2006-01-01

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

CS Informativeness Governs CS-US Associability

Science.gov (United States)

Ward, Ryan D.; Gallistel, C. R.; Jensen, Greg; Richards, Vanessa L.; Fairhurst, Stephen; Balsam, Peter D

2012-01-01

In a conditioning protocol, the onset of the conditioned stimulus (CS) provides information about when to expect reinforcement (the US). There are two sources of information from the CS in a delay conditioning paradigm in which the CS-US interval is fixed. The first depends on the informativeness, the degree to which CS onset reduces the average expected time to onset of the next US. The second depends only on how precisely a subject can represent a fixed-duration interval (the temporal Weber fraction). In three experiments with mice, we tested the differential impact of these two sources of information on rate of acquisition of conditioned responding (CS-US associability). In Experiment 1, we show that associability (the inverse of trials to acquisition) increases in proportion to informativeness. In Experiment 2, we show that fixing the duration of the US-US interval or the CS-US interval or both has no effect on associability. In Experiment 3, we equated the increase in information produced by varying the C̅/T̅ ratio with the increase produced by fixing the duration of the CS-US interval. Associability increased with increased informativeness, but, as in Experiment 2, fixing the CS-US duration had no effect on associability. These results are consistent with the view that CS-US associability depends on the increased rate of reward signaled by CS onset. The results also provide further evidence that conditioned responding is temporally controlled when it emerges. PMID:22468633

Thermally exfoliated graphite oxide

Science.gov (United States)

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

2011-01-01

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

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

International Nuclear Information System (INIS)

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

1977-01-01

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

Multi-Stage Flotation for the Removal of Ash from Fine Graphite Using Mechanical and Centrifugal Forces

OpenAIRE

Xiangning Bu; Tuantuan Zhang; Yaoli Peng; Guangyuan Xie; Erdong Wu

2018-01-01

Graphite ore collected from Hunan province, south China was characterized by chemical analysis, X-ray diffraction, and optical microscopy. Rougher and multi-stage flotation tests using a mechanical flotation cell and a flotation column containing an additional centrifugal force field were carried out to promote its grade and economic value. In rougher flotation, both the mechanical flotation cell and flotation column reduced the ash content of the graphite ore from 15.43% to 10.8%, while the ...

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

Application of ISRM testing methods to fracture toughness testing of graphite

International Nuclear Information System (INIS)

Hashida, T.; Fukasawa, T.; Takahashi, H.; Ishiyama, S.; Oku, T.

1987-01-01

Fracture toughness measurements of nuclear grade graphites, IG11 and PGX, were made by means of AE technique. Tests were conducted on edge-notched round bend bar, edge-notched short bar and round compact tension specimens. These round-shaped specimens used in this study have been proposed for standard fracture toughness tests of rock as a draft of testing standard of International Society for Rock Mechanics (ISRM). Taking the observed nonlinear deformation behavior into account, J-integral approach was utilized to determine the fracture toughness of the graphites. It is shown that the critical J integral determined by AE technique, J iAE , is independent of specimen geometry. Based on this experimental results, the fracture toughness K IC of the graphites was determined from the J iAE values. K IC value of IG11 was 1.04 MPa√m, and 0.77 MPa√m for PGX respectively. Furthermore, the specimen size effect of the fracture toughness determined by the J-integral/AE method is discussed. (author)

Carbon isotope geothermometry of graphite-bearing marbles from Central Dronning Maud Land, East Antarctica

International Nuclear Information System (INIS)

Wand, U.; Muehle, K.

1988-01-01

In order to estimate the peak metamorphic temperatures in high-grade regional metamorphic marbles from central Dronning Maud Land (East Antarctica), 13 C/ 12 C isotope ratios have been measured for coexisting carbonate and graphite pairs. The δ 13 C values of carbonates (calcite ± dolomite) and graphite vary from -0.1 to +4.6 permill (PDB) and from -3.3 to +1.7 permill, respectively. The isotopic fractionation between carbonate and graphite ranges from 2.9 to 4.0 permill and is similar to the Δ 13 C (carb-gr) values observed in other East Antarctic and non-Antarctic granulite-facies marbles. The metamorphic temperatures calculated using the equation of VALLEY and O'NEIL (1981) for calcite-graphite pairs are predominantly in the range 700 0 - 800 0 C (x n=5 ± s = 730 0 ± 30 0 C) and agree well with metamorphic temperatures derived from mineral chemical studies in this East Antarctic region. (author)

Plant remediation of soil contaminated with 137Cs

International Nuclear Information System (INIS)

Yang Juncheng; Zhang Jianfeng; Zhu Yongyi; Chen Jingjie; Mei Yong; Jiang Huimin

2005-01-01

A pot experiment was conducted to evaluate the bio-remediation of soils contaminated with 137 Cs. The selected plants are Cucurbita moschata Duchesne, Brassica chinensis L, Chloris virgata, Beta oulgaris L. Hongye, Beta oulgaris L. Dongshengye and Beta oulgaris L. The soils samples were taken from the paddy field, 2 km from the Dayawan nuclear power plant and Qinshan nuclear power plant, respectively, and cinnamon soil from the cultivated land in Beijing. The results show that all the employed species of plant have a higher accumulation to 137 Cs with the increased grade of the radioactivity of 137 Cs. A good correlation exist with the coefficient (r 2 ) of 0.9989. When the contaminated radioactivity of 137 Cs is in the same level the uptake of Cucurbita moschata Duchesne, Brassica chinensis L. and Chloris virgata increased with the decrease of pH value ranged 5.22-7.69. The ability of bioremediation in the orders were Chloris virgata, Brassica chinensis L., Beta oulgaris L. Hongye, Cucurbita moschata Duchesne, Beta oulgaris L. and Beta oulgaris L. Dongshengye, according to the comprehensive evaluation of transfer factor, specific activity of plant in dry weight of biomass and total absorption of 137 Cs by the individual plant in the same area. (authors)

On the isotropic elastic constants of graphite nodules in ductile cast iron: Analytical and numerical micromechanical investigations

DEFF Research Database (Denmark)

Andriollo, Tito; Hattel, Jesper

2016-01-01

A comprehensive description of the mechanical behavior of nodules in ductile iron is still missing in the published literature. Nevertheless, experimental evidence exists for the importance of such graphite particles during macroscopic material deformation, especially under compressive loading...... mesoscopic moduli in agreement with Young's modulus and Poisson's ratio recorded for common ferritic ductile iron grades. This suggests that graphite nodules may not be considered isotropic at the microscopic scale, at least from a mechanical viewpoint....

Molecular structures and thermodynamic properties of 12 gaseous cesium-containing species of nuclear safety interest: Cs 2, CsH, CsO, Cs 2O, CsX, and Cs 2X 2 (X = OH, Cl, Br, and I)

Science.gov (United States)

Badawi, Michael; Xerri, Bertrand; Canneaux, Sébastien; Cantrel, Laurent; Louis, Florent

2012-01-01

Ab initio electronic structure calculations at the coupled cluster level with a correction for the triples extrapolated to the complete basis set limit have been made for the estimation of the thermochemical properties of Cs 2, CsH, CsO, Cs 2O, CsX, and Cs 2X 2 (X = OH, Cl, Br, and I). The standard enthalpies of formation and standard molar entropies at 298 K, and the temperature dependence of the heat capacities at constant pressure were evaluated. The calculated thermochemical properties are in good agreement with their literature counterparts. For Cs 2, CsH, CsOH, Cs 2(OH) 2, CsCl, Cs 2Cl 2, CsBr, CsI, and Cs 2I 2, the calculated ΔfH298K∘ values are within chemical accuracy of the most recent experimental values. Based on the excellent agreement observed between our calculated ΔfH298K∘ values and their literature counterparts, the standard enthalpies of formation at 298 K are estimated to be the following: ΔfH298K∘ (CsO) = 17.0 kJ mol -1 and ΔfH298K∘ (Cs 2Br 2) = -575.4 kJ mol -1.

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

Science.gov (United States)

Zhamu, Aruna; Jang, Bor Z.

2014-06-17

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

Monolayer CS as a metal-free photocatalyst with high carrier mobility and tunable band structure: a first-principles study

Science.gov (United States)

Yang, Xiao-Le; Ye, Xiao-Juan; Liu, Chun-Sheng; Yan, Xiao-Hong

2018-02-01

Producing hydrogen fuel using suitable photocatalysts from water splitting is a feasible method to harvest solar energy. A desired photocatalyst is expected to have suitable band gap, moderate band edge position, and high carrier mobility. By employing first-principles calculations, we explore a α-CS monolayer as a metal-free efficient photocatalyst. The α-CS monolayer shows good energetic, dynamic, and thermal stabilities and is insoluble in water, suggesting its experimental practicability. Monolayer and bilayer α-CS present not only appropriate band gaps for visible and ultraviolet light absorption but also moderate band alignments with water redox potentials in pH neutral water. Remarkably, the α-CS monolayer exhibits high (up to 8453.19 cm2 V-1s-1 for hole) and anisotropic carrier mobility, which is favorable to the migration and separation of photogenerated carriers. In addition, monolayer α-CS experiences an interesting semiconductor-metal transition by applying uniaxial strain and external electric field. Moreover, α-CS under certain strain and electric field is still dynamically stable with the absence of imaginary frequencies. Furthermore, we demonstrate that the graphite (0 0 1) surface is a potential substrate for the α-CS growth with the intrinsic properties of α-CS maintaining. Therefore, our results could pave the way for the application of α-CS as a promising photocatalyst.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-01

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

Reinforcement of cement-based matrices with graphite nanomaterials

Science.gov (United States)

Sadiq, Muhammad Maqbool

Cement-based materials offer a desirable balance of compressive strength, moisture resistance, durability, economy and energy-efficiency; their tensile strength, fracture energy and durability in aggressive environments, however, could benefit from further improvements. An option for realizing some of these improvements involves introduction of discrete fibers into concrete. When compared with today's micro-scale (steel, polypropylene, glass, etc.) fibers, graphite nanomaterials (carbon nanotube, nanofiber and graphite nanoplatelet) offer superior geometric, mechanical and physical characteristics. Graphite nanomaterials would realize their reinforcement potential as far as they are thoroughly dispersed within cement-based matrices, and effectively bond to cement hydrates. The research reported herein developed non-covalent and covalent surface modification techniques to improve the dispersion and interfacial interactions of graphite nanomaterials in cement-based matrices with a dense and well graded micro-structure. The most successful approach involved polymer wrapping of nanomaterials for increasing the density of hydrophilic groups on the nanomaterial surface without causing any damage to the their structure. The nanomaterials were characterized using various spectrometry techniques, and SEM (Scanning Electron Microscopy). The graphite nanomaterials were dispersed via selected sonication procedures in the mixing water of the cement-based matrix; conventional mixing and sample preparation techniques were then employed to prepare the cement-based nanocomposite samples, which were subjected to steam curing. Comprehensive engineering and durability characteristics of cement-based nanocomposites were determined and their chemical composition, microstructure and failure mechanisms were also assessed through various spectrometry, thermogravimetry, electron microscopy and elemental analyses. Both functionalized and non-functionalized nanomaterials as well as different

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

Energy Technology Data Exchange (ETDEWEB)

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

1955-07-01

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

Feasibility of monitoring the strength of HTGR core support graphite: Part III

International Nuclear Information System (INIS)

Morgan, W.C.; Davis, T.J.; Thomas, M.T.

1983-02-01

Methods are being developed to monitor, in-situ, the strength changes of graphite core-support components in a High-Temperature Gas-Cooled Reactor (HTGR). The results reported herein pertain to the development of techniques for monitoring the core-support blocks; the PGX graphite used in these studies is the grade used for the core-support blocks of the Fort St. Vrain HTGR, and is coarser-grained than the grades used in our previous investigations. The through-transmission ultrasonic velocity technique, developed for monitoring strength of the core-support posts, is not suitable for use on the core-support blocks. Eddy-current and ultrasonic backscattering techniques have been shown to be capable of measuring the density-depth profile in oxidized PGX and, combined with a correlation of strength versus density, could yield an estimate of the strength-depth profile of in-service HTGR core support blocks. Correlations of strength versus density and other properties, and progress on the development of the eddy-current and ultrasonic backscattering techniques are reported

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

Interactions of hydrogen with graphite at low pressure and elevated temperature

International Nuclear Information System (INIS)

Hoinkis, E.

1991-03-01

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

Molecular CsF 5 and CsF 2 +

KAUST Repository

Rogachev, Andrey Yu.; Miao, Mao-sheng; Merino, Gabriel; Hoffmann, Roald

2015-01-01

D5h star-like CsF5, formally isoelectronic with known XeF5− ion, is computed to be a local minimum on the potential energy surface of CsF5, surrounded by reasonably large activation energies for its exothermic decomposition to CsF+2 F2, or to CsF3 (three isomeric forms)+F2, or for rearrangement to a significantly more stable isomer, a classical Cs+ complex of F5−. Similarly the CsF2+ ion is computed to be metastable in two isomeric forms. In the more symmetrical structures of these molecules there is definite involvement in bonding of the formally core 5p levels of Cs.

Molecular CsF 5 and CsF 2 +

KAUST Repository

Rogachev, Andrey Yu.

2015-06-03

D5h star-like CsF5, formally isoelectronic with known XeF5− ion, is computed to be a local minimum on the potential energy surface of CsF5, surrounded by reasonably large activation energies for its exothermic decomposition to CsF+2 F2, or to CsF3 (three isomeric forms)+F2, or for rearrangement to a significantly more stable isomer, a classical Cs+ complex of F5−. Similarly the CsF2+ ion is computed to be metastable in two isomeric forms. In the more symmetrical structures of these molecules there is definite involvement in bonding of the formally core 5p levels of Cs.

Behavior of LASL-made graphite, ZrC, and ZrC-containing coated particles in irradiation tests HT-28 and HT-29

International Nuclear Information System (INIS)

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

1976-01-01

Three types of materials, extruded graphite, hot-pressed ZrC, and particles with ZrC coatings, were irradiated in ORNL High Fluence Isotope Reactor Irradiation tests HT-28 and HT-29. The ZrC seemed unaffected. The graphite changed in dimensions, x-ray diffraction parameters, and thermal conductivity. The four types of coated particles tested all resisted the irradiation well, except one set of particles with double-graded C-ZrC-C coats. Overall, the results were considered encouraging for use of ZrC and extruded graphite fuel matrices. 16 fig

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

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)

THE EFFECT OF APPLIED STRESS ON THE GRAPHITIZATION OF PYROLYTIC GRAPHITE

Energy Technology Data Exchange (ETDEWEB)

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

1963-06-15

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

Neutron scattering investigation of layer-bending modes in alkali-metal--graphite intercalation compounds

International Nuclear Information System (INIS)

Zabel, H.; Kamitakahara, W.A.; Nicklow, R.M.

1982-01-01

Phonon dispersion curves for low-frequency transverse modes propagating in the basal plane have been measured in the alkali-metal--graphite intercalation compounds KC 8 , CsC 8 , KC 24 , and RbC 24 by means of neutron spectroscopy. The acoustic branches show an almost quadratic dispersion relation at small q, characteristic of strongly layered materials. The optical branches of stage-1 compounds can be classified as either graphitelike branches showing dispersion, or as almost dispersionless alkali-metal-like modes. Macroscopic shear constants C 44 and layer-bending moduli have been obtained for the intercalation compounds by analyzing the data in terms of a simple semicontinuum model. In stage-2 compounds, a dramatic softening of the shear constant by about a factor of 8 compared with pure graphite has been observed. Low-temperature results on KC 24 indicate the opening of a frequency gap near the alkali-metal Brillouin-zone boundary, possibly due to the formation of the alkali-metal superstructure

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

Bridged graphite oxide materials

Science.gov (United States)

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

2010-01-01

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

Performance study on a low-temperature absorption–compression cascade refrigeration system driven by low-grade heat

International Nuclear Information System (INIS)

Xu, Yingjie; Chen, Guangming; Wang, Qin; Han, Xiaohong; Jiang, Ning; Deng, Shiming

2016-01-01

Highlights: • An absorption–compression system for low-temperature is developed and analyzed. • Cooling capacity, compression power, and discharge temperature are all improved. • At −170 °C, giving 200 W low-grade cooling capacity, COP increases by 28.6%. • Simulation results are verified experimentally, showing good agreement. - Abstract: This paper presents a performance study on a low-temperature absorption–compression cascade refrigeration system (LACRS), which consists of an absorption subsystem (AS) and a vapor compression auto-cascade subsystem (CS). In the system, low-grade heat of AS is used to subcool the CS, which can obtain cold energy at −170 °C. A simulation study is carried out to investigate the effects of evaporating temperature and low-grade cooling capacity on system performance. The study results show that as low-grade cooling capacity from the AS is provided to the CS, high-grade cooling capacity increases, compressor power consumption decreases, and the COP of the CS therefore increases. Comparing with compression auto-cascade cycle, the largest COP improvement of LACRS is about 38%. The model is verified by experimental data. An additional high-grade cooling capacity is obtained experimentally at −170 °C. The study results presented in this paper not only demonstrate the excellent performance of the LACRS, but also provide important guidance to further system design, and practical application.

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Statistical analysis of surface roughness of machined graphite by means of CNC milling

Directory of Open Access Journals (Sweden)

Orquídea Sánchez López

2016-09-01

Full Text Available The aim of this research is to analyze the influence of cutting speed, feed rate and cutting depth on the surface finish of grade GSP-70 graphite specimens for use in electrical discharge machining (EDM for material removal by means of Computer Numerical Control (CNC milling with low-speed machining (LSM. A two-level factorial design for each of the three established factors was used for the statistical analysis. The analysis of variance (ANOVA indicates that cutting speed and feed rate are the two most significant factors with regard to the roughness obtained with grade GSP-70 graphite by means of CNC milling. A second order regression analysis was also conducted to estimate the roughness average (Ra in terms of the cutting speed, feed rate and cutting depth. Finally, the comparison between predicted roughness by means of a second order regression model and the roughness obtained by machined specimens considering the combinations of low and high levels of roughness is also presented.

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

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

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

Microstructural characteristics, mechanical and wear behaviour of aluminium matrix hybrid composites reinforced with alumina, rice husk ash and graphite

Directory of Open Access Journals (Sweden)

Kenneth Kanayo Alaneme

2015-09-01

Full Text Available The microstructural characteristics, mechanical and wear behaviour of Aluminium matrix hybrid composites reinforced with alumina, rice husk ash (RHA and graphite were investigated. Alumina, RHA and graphite mixed in varied weight ratios were utilized to prepare 10 wt% hybrid reinforced Al-Mg-Si alloy based composites using two-step stir casting. Hardness, tensile properties, scanning electron microscopy, and wear tests were used to characterize the composites produced. The results show that Hardness decreases with increase in the weight ratio of RHA and graphite in the composites; and with RHA content greater than 50%, the effect of graphite on the hardness becomes less significant. The tensile strength for the composites containing o.5wt% graphite and up to 50% RHA was observed to be higher than that of the composites without graphite. The toughness values for the composites containing 0.5wt% graphite were in all cases higher than that of the composites without graphite. The % Elongation for all composites produced was within the range of 10–13% and the values were invariant to the RHA and graphite content. The tensile fracture surface morphology in all the composites produced was identical characterized with the presence of reinforcing particles housed in ductile dimples. The composites without graphite exhibited greater wear susceptibility in comparison to the composite grades containing graphite. However the wear resistance decreased with increase in the graphite content from 0.5 to 1.5 wt%.

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

Science.gov (United States)

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

2017-02-01

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

Graphite selection for the PBMR reflector

International Nuclear Information System (INIS)

Marsden, B.J.; Preston, S.D.

2000-01-01

A high temperature, direct cycle gas turbine, graphite moderated, helium cooled, pebble-bed reactor (PBMR) is being designed and constructed in South Africa. One of the major components in the PBMR is the graphite reflector, which must be designed to last thirty-five full power years. Fast neutron irradiation changes the dimensions and material properties of reactor graphite, thus for design purposes a suitable graphite database is required. Data on the effect of irradiation on nuclear graphites has been gathered for many years, at considerable financial cost, but unfortunately these graphites are no longer available due to rationalization of the graphite industry and loss of key graphite coke supplies. However, it is possible, using un-irradiated graphite materials properties and knowledge of the particular graphite microstructure, to determine the probable irradiation behaviour. Three types of nuclear graphites are currently being considered for the PBMR reflector: an isostatically moulded, fine grained, high strength graphite and two extruded medium grained graphites of moderately high strength. Although there is some irradiation data available for these graphites, the data does not cover the temperature and dose range required for the PBMR. The available graphites have been examined to determine their microstructure and some of the key material properties are presented. (authors)

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

Science.gov (United States)

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

2010-11-02

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

A study of the coefficient of thermal expansion of nuclear graphites

International Nuclear Information System (INIS)

Hacker, P.J.

2001-02-01

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

Preparation and evaluation of coal extracts as precursors for carbon and graphite products

Energy Technology Data Exchange (ETDEWEB)

Zondlo, J.W.; Stiller, A.W.; Stansberry, P.G. [West Virginia Univ., Morgantown, WV (United States)] [and others

1996-08-01

A coal extraction process coupled with coal hydrotreatment has been shown capable of producing suitable precursors for a variety of commercially important carbon and graphite products. The N-methylpyrolidone (NMP) extracts of hydrotreated coals have been analytically and chemically characterized and shown to have properties acceptable for use as binder and impregnation pitch. Mesophase formation studies have demonstrated their capability for producing both needle and anode grade coke as well as precursors for mesophase pitch fibers. A graphite artifact has been produced using a coal extract as a binder and coke derived from the extract as a filler. Further evaluation of the extract materials is being carried out by industrial members of the Carbon Products Consortium.

Immobilisation of 137Cs and 90Sr in hot-pressed clinoptilolite

International Nuclear Information System (INIS)

Briggs, A.; Jones, D.V.C.; Cole, G.B.; Valentine, T.M.; Preston, R.F.; Hawes, R.W.M.; Fones, M.D.

1984-07-01

It is shown that clinoptilolites from Nevada and Mudhills, California, can be vitrified by hot-pressing at 1000 to 1050 deg C for 30 min under 23 MPa pressure in graphite dies, or for 10 min under 3.5 MPa pressure in stainless steel bellows-shaped containers. Activity measurements before and after hot-pressing show that the radionuclides 137 Cs and 90 Sr are retained almost completely during the process. The hot-pressed material is shown to be extremely resistant to leaching of the nuclides, either in Soxhlet or long-term static leach tests. Gamma-irradiation increases their leachability; that of 90 Sr more than that of 137 Cs, but even so, the leach rates of both remain very low. X-ray diffraction spectra taken before and after heat-treatment for 10 days at temperatures up to 1000 deg C show no deleterious effects, and hydrothermal treatments at temperatures up to 90 deg C for 100 days, even in the presence of sodium ions in solution, are shown not to cause any changes detectable by X-ray diffraction. (author)

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.

Reconstruction of 137Cs signal in cuba using 7be as a Tracer of vertical transport processes in the Atmosphere

International Nuclear Information System (INIS)

Alonso Hernandez, C.M.; Cartas Aguila, H.; Diaz Asencio, M.; Munnoz Caravaca, A.

2001-01-01

The goal of this study is to estimate the 137Cs signal in Cuba using the relation 137Cs/7Be for Miami and Cienfuegos City's obtained for four year and data base of 137Cs concentration in aerosols for Miami since 1957. The temporal behaviour of 137Cs and 7Be in aerosols in the Central and Southern region of Cuba are shown too. Sampling of aerosols was carried out in the Environmental Study Centre (22 grade 03'N, 80 grade 29'W) facilities during June/1994 - December/1998 in Cienfuegos Province. Aerosols were collected by an integral method using the facility Primus I, with 0.65 m2 collection area, 1200 m3h-1 air flux during 7 days collection time. Petrianov filter FPP-15-1.5 was used (efficiency collection of 99% to aerosols 0.3-1.5 mm diameter). After collection, the filters were ashed at 350 grade for 72 hours, put in appropriate geometry and analysed by gamma spectrometry. 137Cs mean concentration in aerosols was 1.09 mBqm-3 and the range of variations was 0.18-3.24 mBqm-3 ; while the 7Be was 4.10 mBqm-3 with 0.96-10.46 mBqm-3 variation range. These results allow to establish a relationship between 137Cs concentrations in aerosols from Cienfuegos and Miami in the same period, normalising in both cases with 7Be values; this normalisation allows to discriminate the variations due to local meteorological conditions and latitudinal position

A graphite nanoeraser

DEFF Research Database (Denmark)

Liu, Ze; Bøggild, Peter; Yang, Jia-rui

2011-01-01

We present here a method for cleaning intermediate-size (up to 50 nm) contamination from highly oriented pyrolytic graphite and graphene. Electron-beam-induced deposition of carbonaceous material on graphene and graphite surfaces inside a scanning electron microscope, which is difficult to remove...... by conventional techniques, can be removed by direct mechanical wiping using a graphite nanoeraser, thus drastically reducing the amount of contamination. We discuss potential applications of this cleaning procedure....

Performance of AC/graphite capacitors at high weight ratios of AC/graphite

Energy Technology Data Exchange (ETDEWEB)

Wang, Hongyu [IM and T Ltd., Advanced Research Center, Saga University, 1341 Yoga-machi, Saga 840-0047 (Japan); Yoshio, Masaki [Advanced Research Center, Department of Applied Chemistry, Saga University, 1341 Yoga-machi, Saga 840-0047 (Japan)

2008-03-01

The effect of negative to positive electrode materials' weight ratio on the electrochemical performance of both activated carbon (AC)/AC and AC/graphite capacitors has been investigated, especially in the terms of capacity and cycle-ability. The limited capacity charge mode has been proposed to improve the cycle performance of AC/graphite capacitors at high weight ratios of AC/graphite. (author)

Progress in radioactive graphite waste management

International Nuclear Information System (INIS)

2010-07-01

Radioactive graphite constitutes a major waste stream which arises during the decommissioning of certain types of nuclear installations. Worldwide, a total of around 250 000 tonnes of radioactive graphite, comprising graphite moderators and reflectors, will require management solutions in the coming years. 14 C is the radionuclide of greatest concern in nuclear graphite; it arises principally through the interaction of reactor neutrons with nitrogen, which is present in graphite as an impurity or in the reactor coolant or cover gas. 3 H is created by the reactions of neutrons with 6 Li impurities in graphite as well as in fission of the fuel. 36 Cl is generated in the neutron activation of chlorine impurities in graphite. Problems in the radioactive waste management of graphite arise mainly because of the large volumes requiring disposal, the long half-lives of the main radionuclides involved and the specific properties of graphite - such as stored Wigner energy, graphite dust explosibility and the potential for radioactive gases to be released. Various options for the management of radioactive graphite have been studied but a generally accepted approach for its conditioning and disposal does not yet exist. Different solutions may be appropriate in different cases. In most of the countries with radioactive graphite to manage, little progress has been made to date in respect of the disposal of this material. Only in France has there been specific thinking about a dedicated graphite waste-disposal facility (within ANDRA): other major producers of graphite waste (UK and the countries of the former Soviet Union) are either thinking in terms of repository disposal or have no developed plans. A conference entitled 'Solutions for Graphite Waste: a Contribution to the Accelerated Decommissioning of Graphite Moderated Nuclear Reactors' was held at the University of Manchester 21-23 March 2007 in order to stimulate progress in radioactive graphite waste management

Multi-Stage Flotation for the Removal of Ash from Fine Graphite Using Mechanical and Centrifugal Forces

Directory of Open Access Journals (Sweden)

Xiangning Bu

2018-01-01

Full Text Available Graphite ore collected from Hunan province, south China was characterized by chemical analysis, X-ray diffraction, and optical microscopy. Rougher and multi-stage flotation tests using a mechanical flotation cell and a flotation column containing an additional centrifugal force field were carried out to promote its grade and economic value. In rougher flotation, both the mechanical flotation cell and flotation column reduced the ash content of the graphite ore from 15.43% to 10.8%, while the yield of the flotation column (91.41% was much higher than that of the mechanical flotation cell (50%. In the presence of hydrophobic graphite, the seriously entrained gangue restricted further improvement in the quality and economic value of the graphite ore. Therefore, multi-stage flotation circuits were employed to diminish this entrainment. Multi-stage flotation circuits using the two flotation devices further decreased the ash content of the graphite ore to ~8%, while the yield when using the flotation column was much higher than that obtained from the mechanical flotation cell employed. On the other hand, the ash removal efficiency of the flotation column was 3.82-fold higher than that observed for the mechanical flotation cell. The Cleaner 3 flotation circuit using the flotation column decreased the ash content in graphite from 15.43% to 7.97% with a yield of 77.53%.

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

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.

Graphite in Science and Nuclear Technique

OpenAIRE

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

2013-01-01

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

Management of UKAEA graphite liabilities

International Nuclear Information System (INIS)

Wise, M.

2001-01-01

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

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

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)

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)

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

Phonon scattering in graphite

International Nuclear Information System (INIS)

Wagner, P.

1976-04-01

Effects on graphite thermal conductivities due to controlled alterations of the graphite structure by impurity addition, porosity, and neutron irradiation are shown to be consistent with the phonon-scattering formulation 1/l = Σ/sub i equals 1/sup/n/ 1/l/sub i/. Observed temperature effects on these doped and irradiated graphites are also explained by this mechanism

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

NARCIS (Netherlands)

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

1969-01-01

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

Electrochemical treatment of graphite

Energy Technology Data Exchange (ETDEWEB)

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

1983-01-01

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

Effect of Graphite Electrode to Surface’s Characteristic of EDM

Directory of Open Access Journals (Sweden)

Muttamara Apiwat

2016-01-01

Full Text Available Electrical discharge machining process (EDM is a process for removing material by the thermal of electrical discharge. Some of the melted and all of the evaporated material is then quenched and flushed away by dielectric liquid and the remaining melt recast on the finished surface. The recast layer is called as white layer. Beneath the recast layer, a heat affected zone is formed. The quality of an EDM product is usually evaluated in terms of its surface integrity, which is characterized by the surface roughness, existence of surface cracks and residual stresses. This paper presents a study of surface’s characteristics by EDM in de-ionized water due to decarbonisation. The machining tests were conducted on mild steel JIS grade SS400 with copper and graphite electrodes. The workpiece surfaces are analyzed by scanning electron microscope and XRD technique. The carbon transfers from graphite electrode to the white layer relating to martensitic phrase of recast layer.

Pattern of Tumor Shrinkage during Neoadjuvant Chemotherapy Is Associated with Prognosis in Low-Grade Luminal Early Breast Cancer.

Science.gov (United States)

Fukada, Ippei; Araki, Kazuhiro; Kobayashi, Kokoro; Shibayama, Tomoko; Takahashi, Shunji; Gomi, Naoya; Kokubu, Yumi; Oikado, Katsunori; Horii, Rie; Akiyama, Futoshi; Iwase, Takuji; Ohno, Shinji; Hatake, Kiyohiko; Sata, Naohiro; Ito, Yoshinori

2018-01-01

Purpose To evaluate the association between tumor shrinkage patterns shown with magnetic resonance (MR) imaging during neoadjuvant chemotherapy (NAC) and prognosis in patients with low-grade luminal breast cancer. Materials and Methods This retrospective study was approved by the institutional review board and informed consent was obtained from all subjects. The low-grade luminal breast cancer was defined as hormone receptor-positive and human epidermal growth factor receptor 2-negative with nuclear grades 1 or 2. The patterns of tumor shrinkage as revealed at MR imaging were categorized into two types: concentric shrinkage (CS) and non-CS. Among 854 patients who had received NAC in a single institution from January 2000 to December 2009, 183 patients with low-grade luminal breast cancer were retrospectively evaluated for the development set. Another data set from 292 patients who had received NAC in the same institution between January 2010 and December 2012 was used for the validation set. Among these 292 patients, 121 patients with low-grade luminal breast cancer were retrospectively evaluated. Results In the development set, the median observation period was 67.9 months. Recurrence was observed in 31 patients, and 16 deaths were related to breast cancer. There were statistically significant differences in both the disease-free survival (DFS) and overall survival (OS) rates between patterns of tumor shrinkage (P breast cancer. DFS rate was significantly longer in patients with the CS pattern (72.8 months; 95% confidence interval [CI]: 69.9, 75.6 months) than in those with the non-CS pattern (56.0 months; 95% CI: 49.1, 62.9 months; P ≤ .001). The CS pattern was associated with an excellent prognosis (median OS, 80.6 months; 95% CI: 79.3, 81.8 months vs 65.0 months; 95% CI: 60.1, 69.8 months; P = .004). Multivariate analysis demonstrated that the CS pattern had the only significant independent association with DFS (P = .007) and OS (P = .037) rates. Conclusion

Graphite-graphite oxide composite electrode for vanadium redox flow battery

International Nuclear Information System (INIS)

Li Wenyue; Liu Jianguo; Yan Chuanwei

2011-01-01

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

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.

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).

Graphite in Science and Nuclear Technology

OpenAIRE

Zhmurikov, Evgenij

2015-01-01

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

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

International Nuclear Information System (INIS)

Matsuo, Hideto

1991-06-01

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

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.

Analysis of Wigner energy release process in graphite stack of shut-down uranium-graphite reactor

OpenAIRE

Bespala, E. V.; Pavliuk, A. O.; Kotlyarevskiy, S. G.

2015-01-01

Data, which finding during thermal differential analysis of sampled irradiated graphite are presented. Results of computational modeling of Winger energy release process from irradiated graphite staking are demonstrated. It's shown, that spontaneous combustion of graphite possible only in adiabatic case.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-02-20

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

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

International Nuclear Information System (INIS)

Anfimov, S.S.

2001-01-01

In this paper an radioactive waste processing of graphite from graphite moderated nuclear reactors at its decommissioning is discussed. Methods of processing of irradiated graphite are presented. It can be concluded that advanced methods for graphite radioactive waste handling are available nowadays. Implementation of these methods will allow to enhance environmental safety of nuclear power that will benefit its progress in the future

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.

Antibodies derived from an enterotoxigenic Escherichia coli (ETEC) adhesin tip MEFA (multiepitope fusion antigen) against adherence of nine ETEC adhesins: CFA/I, CS1, CS2, CS3, CS4, CS5, CS6, CS21 and EtpA.

Science.gov (United States)

Nandre, Rahul M; Ruan, Xiaosai; Duan, Qiangde; Sack, David A; Zhang, Weiping

2016-06-30

Diarrhea continues to be a leading cause of death in children younger than 5 years in developing countries. Enterotoxigenic Escherichia coli (ETEC) is a leading bacterial cause of children's diarrhea and travelers' diarrhea. ETEC bacteria initiate diarrheal disease by attaching to host receptors at epithelial cells and colonizing in small intestine. Therefore, preventing ETEC attachment has been considered the first line of defense against ETEC diarrhea. However, developing vaccines effectively against ETEC bacterial attachment encounters challenge because ETEC strains produce over 23 immunologically heterogeneous adhesins. In this study, we applied MEFA (multiepitope fusion antigen) approach to integrate epitopes from adhesin tips or adhesive subunits of CFA/I, CS1, CS2, CS3, CS4, CS5, CS6, CS21 and EtpA adhesins and to construct an adhesin tip MEFA peptide. We then examined immunogenicity of this tip MEFA in mouse immunization, and assessed potential application of this tip MEFA for ETEC vaccine development. Data showed that mice intraperitoneally immunized with this adhesin tip MEFA developed IgG antibody responses to all nine ETEC adhesins. Moreover, ETEC and E. coli bacteria expressing these nine adhesins, after incubation with serum of the immunized mice, exhibited significant reduction in attachment to Caco-2 cells. These results indicated that anti-adhesin antibodies induced by this adhesin tip MEFA blocked adherence of the most important ETEC adhesins, suggesting this multivalent tip MEFA may be useful for developing a broadly protective anti-adhesin vaccine against ETEC diarrhea. Copyright © 2016 Elsevier Ltd. All rights reserved.

Molecular CsF{sub 5} and CsF{sub 2}{sup +}

Energy Technology Data Exchange (ETDEWEB)

Rogachev, Andrey Yu. [Illinois Institute of Technology, IL (United States). Dept. of Biological and Chemical Sciences; Miao, Mao-sheng [California State Univ., Northridge, CA (United States). Dept. of Chemistry and Biochemistry; Beijing Computational Science Research Center (China); Merino, Gabriel [Centro de Investigacion y de Estudios Avanzados, Unidad Merida (Mexico). Dept. de Fisica Aplicada; Hoffmann, Roald [Cornell Univ., Ithaca, NY (United States). Dept. of Chemistry and Chemical Biology

2015-07-06

D{sub 5h} star-like CsF{sub 5}, formally isoelectronic with known XeF{sub 5}{sup -} ion, is computed to be a local minimum on the potential energy surface of CsF{sub 5}, surrounded by reasonably large activation energies for its exothermic decomposition to CsF + 2F{sub 2}, or to CsF{sub 3} (three isomeric forms) + F{sub 2}, or for rearrangement to a significantly more stable isomer, a classical Cs{sup +} complex of F{sub 5}{sup -}. Similarly the CsF{sub 2}{sup +} ion is computed to be metastable in two isomeric forms. In the more symmetrical structures of these molecules there is definite involvement in bonding of the formally core 5p levels of Cs.

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

Science.gov (United States)

Jones, Abbie N.; Marsden, Barry J.

2017-01-01

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

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

Directory of Open Access Journals (Sweden)

Alex Theodosiou

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

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.

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

Science.gov (United States)

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

2016-09-01

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

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

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

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)

Analysis of 137Cs radionuclides activities in Cs-zeolite using gamma spectrometry

International Nuclear Information System (INIS)

Noviarty; Ginting, Aslina Br; Anggraini, Dian; Rosika K

2013-01-01

137 Cs Radionuclide activity analysis has been carried out. The objective is to determine the activity of the 137 Cs radionuclide in Cs-zeolite are packaged in the form of point source. Analysis of 137 Cs Radionuclide activities in Cs zeolite samples was determined by measuring intensity of the isotope 137 Cs gamma energy at 661.7 keV use-y spectrometer. Before measurement the sample, was first carried out measurements of 137 Cs radionuclide in certified point standards from Amersham, to determine the efficiency value. Result the standard sample measurement obtained the efficiency value of 43.98%. Efficiency values obtained are used in the calculation of sample activity. On the measurement of the intensity of the sample obtained results dose rate 196.4537 cps with a standard deviation of 0.5274. By using standard measurement efficiency values obtained by the calculation of the average activity of the radionuclide 137 Cs in Cs-zeolite 524.9082 Bq. Deviation measurements were below 5% (0.27% ) so that the analysis of the activity of radionuclide 137 Cs in Cs-zeolite samples using gamma spectrometer can be accepted with a 95% confidence level. (author)

Recompressed exfoliated graphite articles

Science.gov (United States)

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

2013-08-06

This invention provides an electrically conductive, less anisotropic, recompressed exfoliated graphite article comprising a mixture of (a) expanded or exfoliated graphite flakes; and (b) particles of non-expandable graphite or carbon, wherein the non-expandable graphite or carbon particles are in the amount of between about 3% and about 70% by weight based on the total weight of the particles and the expanded graphite flakes combined; wherein the mixture is compressed to form the article having an apparent bulk density of from about 0.1 g/cm.sup.3 to about 2.0 g/cm.sup.3. The article exhibits a thickness-direction conductivity typically greater than 50 S/cm, more typically greater than 100 S/cm, and most typically greater than 200 S/cm. The article, when used in a thin foil or sheet form, can be a useful component in a sheet molding compound plate used as a fuel cell separator or flow field plate. The article may also be used as a current collector for a battery, supercapacitor, or any other electrochemical cell.

Metal/graphite-composite materials for fusion device

International Nuclear Information System (INIS)

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

1995-01-01

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

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)

Evaluation of cutting force and surface roughness in high-speed milling of compacted graphite iron

Directory of Open Access Journals (Sweden)

Azlan Suhaimi Mohd

2017-01-01

Full Text Available Compacted Graphite Iron, (CGI is known to have outstanding mechanical strength and weight-to-strength ratio as compared to conventional grey cast iron, (CI. The outstanding characteristics of CGI is due to its graphite particle shape, which is presented as compacted vermicular particle. The graphite is interconnected with random orientation and round edges, which results in higher mechanical strength. Whereas, graphite in the CI consists of a smooth-surfaced flakes that easily propagates cracks which results in weaker and brittle properties as compared to CGI. Owing to its improved properties, CGI is considered as the best candidate material in substituting grey cast iron that has been used in engine block applications for years. However, the smooth implementation of replacing CI with CGI has been hindered due to the poor machinability of CGI especially at high cutting speed. The tool life is decreased by 20 times when comparing CGI with CI under the same cutting condition. This study investigates the effect of using cryogenic cooling and minimum quantity lubrication (MQL during high-speed milling of CGI (grade 450. Results showed that, the combination of internal cryogenic cooling and enhanced MQL improved the tool life, cutting force and surface quality as compared to the conventional flood coolant strategy during high-speed milling of CGI.

Characterization of Ignalina NPP RBMK Reactors Graphite

International Nuclear Information System (INIS)

Hacker, P.J.; Neighbour, G.B.; Levinskas, R.; Milcius, D.

2001-01-01

The paper concentrates on the investigations of the initial physical properties of graphite used in production of graphite bricks of Ignalina NPP. These graphite bricks are used as nuclear moderator and major core structural components. Graphite bulk density is calculated by mensuration, pore volumes are measured by investigation of helium gas penetration in graphite pore network, the Young's modulus is determined using an ultrasonic time of flight method, the coefficient of thermal expansion is determined using a Netzsch dilatometer 402C, the fractured and machined graphite surfaces are studied using SEM, impurities are investigated qualitatively by EDAX, the degree of graphitization of the material is tested using X-ray diffraction. (author)

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Bromine intercalated graphite for lightweight composite conductors

KAUST Repository

Amassian, Aram; Patole, Archana

2017-01-01

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

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.

Metal/graphite - composites in fusion engineering

International Nuclear Information System (INIS)

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

1995-01-01

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

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)

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.

'Wash-out' of Cs-134 and Cs-137 from river sediment; 'Ispiranja' Cs-134 i Cs-137 iz recnog sedimenta

Energy Technology Data Exchange (ETDEWEB)

Skrbic, Z; Conkic, Lj; Bikit, I; Veskovic, M; Slivka, J; Marinkov, L [Institut za Fiziku, Novi Sad Univ. (Yugoslavia)

1988-07-01

Natural elimination and 'wash out' period of the Cs-134 and Cs-137 from the river sediment has been investigated. Obtained results suggest the possibility to describe these processes by exponential low and determination of the corresponding half lives. (author)

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)

Computational prediction of dust production in graphite moderated pebble bed reactors

Science.gov (United States)

Rostamian, Maziar

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

Porous (Swiss-Cheese Graphite

Directory of Open Access Journals (Sweden)

Joseph P. Abrahamson

2018-05-01

Full Text Available Porous graphite was prepared without the use of template by rapidly heating the carbonization products from mixtures of anthracene, fluorene, and pyrene with a CO2 laser. Rapid CO2 laser heating at a rate of 1.8 × 106 °C/s vaporizes out the fluorene-pyrene derived pitch while annealing the anthracene coke. The resulting structure is that of graphite with 100 nm spherical pores. The graphitizablity of the porous material is the same as pure anthracene coke. Transmission electron microscopy revealed that the interfaces between graphitic layers and the pore walls are unimpeded. Traditional furnace annealing does not result in the porous structure as the heating rates are too slow to vaporize out the pitch, thereby illustrating the advantage of fast thermal processing. The resultant porous graphite was prelithiated and used as an anode in lithium ion capacitors. The porous graphite when lithiated had a specific capacity of 200 mAh/g at 100 mA/g. The assembled lithium ion capacitor demonstrated an energy density as high as 75 Wh/kg when cycled between 2.2 V and 4.2 V.

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

International Nuclear Information System (INIS)

Gesh, Christopher J.

2004-01-01

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

Uptake and distribution of 137Cs, stable Cs and K in rice plants

International Nuclear Information System (INIS)

Tsukada, Hirofumi; Hasegawa, Hidenao

2003-01-01

The uptake and distributions of 137 Cs, stable Cs and K were determined for rice plant components, including polished rice, rice bran, hulls, leaves, stems, and roots. The distribution of 137 Cs in polished rice and rice bran was similar to that of stable Cs, while that of K was different. The concentration ratios of Cs/K in leaves increased in older leaf blade positions, which meant that the translocation rate of stable Cs, was slower than that of K. At harvest the dry weight of polished rice accounted for 34% of the entire rice plant, while the distributions of stable Cs in the polished rice and the non-edible parts were 7 and 93%, respectively. These findings suggest that the transfer and distribution of stable Cs in rice plants are different from those of K, and the behavior of stable Cs provides a useful analogue in predicting the fate of 137 Cs in an agricultural environment. (author)

Graphite paper-based bipolar electrode electrochemiluminescence sensing platform.

Science.gov (United States)

Zhang, Xin; Ding, Shou-Nian

2017-08-15

In this work, aiming at the construction of a disposable, wireless, low-cost and sensitive system for bioassay, we report a closed bipolar electrode electrochemiluminescence (BPE-ECL) sensing platform based on graphite paper as BPE for the first time. Graphite paper is qualified as BPE due to its unique properties such as excellent electrical conductivity, uniform composition and ease of use. This simple BPE-ECL device was applied to the quantitative analysis of oxidant (H 2 O 2 ) and biomarker (CEA) respectively, according to the principle of BPE sensing-charge balance. For the H 2 O 2 analysis, Pt NPs were electrodeposited onto the cathode through a bipolar electrodeposition approach to promote the sensing performance. As a result, this BPE-ECL device exhibited a wide linear range of 0.001-15mM with a low detection limit of 0.5µM (S/N=3) for H 2 O 2 determination. For the determination of CEA, chitosan-multi-walled carbon nanotubes (CS-MWCNTs) were employed to supply a hydrophilic interface for immobilizing primary antibody (Ab 1 ); and Au@Pt nanostructures were conjugated with secondary antibody (Ab 2 ) as catalysts for H 2 O 2 reduction. Under the optimal conditions, the BPE-ECL immunodevice showed a wide linear range of 0.01-60ngmL -1 with a detection limit of 5.0pgmL -1 for CEA. Furthermore, it also displayed satisfactory selectivity, excellent stability and good reproducibility. The developed method opened a new avenue to clinical bioassay. Copyright © 2017 Elsevier B.V. All rights reserved.

Metal/graphite - composites in fusion engineering

International Nuclear Information System (INIS)

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

1989-01-01

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

135Cs activity and 135Cs/137Cs atom ratio in environmental samples before and after the Fukushima Daiichi Nuclear Power Plant accident.

Science.gov (United States)

Yang, Guosheng; Tazoe, Hirofumi; Yamada, Masatoshi

2016-04-07

(135)Cs/(137)Cs is a potential tracer for radiocesium source identification. However, due to the challenge to measure (135)Cs, there were no (135)Cs data available for Japanese environmental samples before the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident. It was only 3 years after the accident that limited (135)Cs values could be measured in heavily contaminated environmental samples. In the present study, activities of (134)Cs, (135)Cs, and (137)Cs, along with their ratios in 67 soil and plant samples heavily and lightly contaminated by the FDNPP accident were measured by combining γ spectrometry with ICP-MS/MS. The arithmetic means of the (134)Cs/(137)Cs activity ratio (1.033 ± 0.006) and (135)Cs/(137)Cs atom ratio (0.334 ± 0.005) (decay corrected to March 11, 2011), from old leaves of plants collected immediately after the FDNPP accident, were confirmed to represent the FDNPP derived radiocesium signature. Subsequently, for the first time, trace (135)Cs amounts before the FDNPP accident were deduced according to the contribution of global and FDNPP accident-derived fallout. Apart from two soil samples with a tiny global fallout contribution, contributions of global fallout radiocesium in other soil samples were observed to be 0.338%-52.6%. The obtained (135)Cs/(137)Cs database will be useful for its application as a geochemical tracer in the future.

Method of Joining Graphite Fibers to a Substrate

Science.gov (United States)

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

2014-01-01

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

Activity standardization of 134Cs and 137Cs

International Nuclear Information System (INIS)

Sochorová, Jana; Auerbach, Pavel

2014-01-01

The paper presents the results from a primary standardization of 137 Cs using two independent methods – efficiency tracing using PC-NaI coincidence and the TDCR method. The nuclides 60 Co and 134 Cs were used as the tracers. Primary standardization of the 134 Cs is also discussed. The efficiency extrapolation was carried out by measuring samples of varying mass and using the wet extrapolation method. The results obtained are in good agreement; the differences did not exceed 0.5%. The advantages, pitfalls and also possibilities for improvement of the procedures are discussed

Determination of Cs-134 and Cs-137 rain water samples

International Nuclear Information System (INIS)

Lima, M.F.; Mazzilli, B.

1988-01-01

In order to setting an environmental monitoring program at IPEN, was developed a fast and simple methodology for concentration of Cs-134 and Cs-137 in rain water. This procedure consists in the precipitation of cesium and others cathions of its family (NH 4 + , K + and Rb + ) by ammonium molybdophosphate. The measures of the desintegration rates of Cs-134 and Cs-137 was done by gamma spectrometry in a Ge(Li) detector. After setting up the ideal experimental conditions, the procedure was used to analyze four samples of rain water. (author) [pt

Mesostructure of graphite composite and its lifetime

OpenAIRE

Zhmurikov, Evgenij

2015-01-01

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

Tungsten Deposition on Graphite using Plasma Enhanced Chemical Vapour Deposition

International Nuclear Information System (INIS)

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

2016-01-01

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

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

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

International Nuclear Information System (INIS)

Anfimov, S.S.

2000-01-01

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

Determination of {sup 135}Cs and {sup 137}Cs in environmental samples: A review

Energy Technology Data Exchange (ETDEWEB)

Russell, B.C., E-mail: ben.russell@npl.co.uk [GAU-Radioanalytical, Ocean and Earth Science, University of Southampton, National Oceanography Centre, Southampton, SO14 3ZH (United Kingdom); National Physical Laboratory, Hampton Road, Teddington, Middlesex, TW11 0LW (United Kingdom); Croudace, Ian W.; Warwick, Phil E. [GAU-Radioanalytical, Ocean and Earth Science, University of Southampton, National Oceanography Centre, Southampton, SO14 3ZH (United Kingdom)

2015-08-26

Radionuclides of caesium are environmentally important since they are formed as significant high yield fission products ({sup 135}Cs and {sup 137}Cs) and activation products ({sup 134}Cs and {sup 136}Cs) during nuclear fission. They originate from a range of nuclear activities such as weapons testing, nuclear reprocessing and nuclear fuel cycle discharges and nuclear accidents. Whilst {sup 137}Cs, {sup 134}Cs and {sup 136}Cs are routinely measurable at high sensitivity by gamma spectrometry, routine detection of long-lived {sup 135}Cs by radiometric methods is challenging. This measurement is, however, important given its significance in long-term nuclear waste storage and disposal. Furthermore, the {sup 135}Cs/{sup 137}Cs ratio varies with reactor, weapon and fuel type, and accurate measurement of this ratio can therefore be used as a forensic tool in identifying the source(s) of nuclear contamination. The shorter-lived activation products {sup 134}Cs and {sup 136}Cs have a limited application but provide useful early information on fuel irradiation history and have importance in health physics. Detection of {sup 135}Cs (and {sup 137}Cs) is achievable by mass spectrometric techniques; most commonly inductively coupled plasma mass spectrometry (ICP-MS), as well as thermal ionisation (TIMS), accelerator (AMS) and resonance ionisation (RIMS) techniques. The critical issues affecting the accuracy and detection limits achievable by this technique are effective removal of barium to eliminate isobaric interferences arising from {sup 135}Ba and {sup 137}Ba, and elimination of peak tailing of stable {sup 133}Cs on {sup 135}Cs. Isobaric interferences can be removed by chemical separation, most commonly ion exchange chromatography, and/or instrumental separation using an ICP-MS equipped with a reaction cell. The removal of the peak tailing interference is dependent on the instrument used for final measurement. This review summarizes and compares the analytical procedures

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.

Progress in radioactive graphite waste management. Additional information

International Nuclear Information System (INIS)

2010-06-01

Radioactive graphite constitutes a major waste stream which arises during the decommissioning of certain types of nuclear installations. Worldwide, a total of around 250 000 tonnes of radioactive graphite, comprising graphite moderators and reflectors, will require management solutions in the coming years. 14 C is the radionuclide of greatest concern in nuclear graphite; it arises principally through the interaction of reactor neutrons with nitrogen, which is present in graphite as an impurity or in the reactor coolant or cover gas. 3 H is created by the reactions of neutrons with 6 Li impurities in graphite as well as in fission of the fuel. 36 Cl is generated in the neutron activation of chlorine impurities in graphite. Problems in the radioactive waste management of graphite arise mainly because of the large volumes requiring disposal, the long half-lives of the main radionuclides involved and the specific properties of graphite - such as stored Wigner energy, graphite dust explosibility and the potential for radioactive gases to be released. Various options for the management of radioactive graphite have been studied but a generally accepted approach for its conditioning and disposal does not yet exist. Different solutions may be appropriate in different cases. In most of the countries with radioactive graphite to manage, little progress has been made to date in respect of the disposal of this material. Only in France has there been specific thinking about a dedicated graphite waste-disposal facility (within ANDRA): other major producers of graphite waste (UK and the countries of the former Soviet Union) are either thinking in terms of repository disposal or have no developed plans. A conference entitled 'Solutions for Graphite Waste: a Contribution to the Accelerated Decommissioning of Graphite Moderated Nuclear Reactors' was held at the University of Manchester 21-23 March 2007 in order to stimulate progress in radioactive graphite waste management

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-01

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

Inhibition of oxidation in nuclear graphite

International Nuclear Information System (INIS)

Winston, Philip L.; Sterbentz, James W.; Windes, William E.

2015-01-01

Graphite is a fundamental material of high-temperature gas-cooled nuclear reactors, providing both structure and neutron moderation. Its high thermal conductivity, chemical inertness, thermal heat capacity, and high thermal structural stability under normal and off-normal conditions contribute to the inherent safety of these reactor designs. One of the primary safety issues for a high-temperature graphite reactor core is the possibility of rapid oxidation of the carbon structure during an off-normal design basis event where an oxidising atmosphere (air ingress) can be introduced to the hot core. Although the current Generation IV high-temperature reactor designs attempt to mitigate any damage caused by a postulated air ingress event, the use of graphite components that inhibit oxidation is a logical step to increase the safety of these reactors. Recent experimental studies of graphite containing between 5.5 and 7 wt% boron carbide (B 4 C) indicate that oxidation is dramatically reduced even at prolonged exposures at temperatures up to 900 deg. C. The proposed addition of B 4 C to graphite components in the nuclear core would necessarily be enriched in B-11 isotope in order to minimise B-10 neutron absorption and graphite swelling. The enriched boron can be added to the graphite during billet fabrication. Experimental oxidation rate results and potential applications for borated graphite in nuclear reactor components will be discussed. (authors)

Promoted Ru on high-surface area graphite for efficient miniaturized production of hydrogen from ammonia

DEFF Research Database (Denmark)

Sørensen, Rasmus Zink; Klerke, Asbjørn; Quaade, Ulrich

2006-01-01

decomposition. The catalytic activities for production of hydrogen from ammonia are determined for different promoters and promoter levels on graphite supported ruthenium catalysts. The reactivity trends of the Ru/C catalysts promoted with Cs and Ba are in excellent agreement with those known from earlier......Promoted Ru/C catalysts for decomposition of ammonia are incorporated into micro-fabricated reactors for the first time. With the reported preparation technique, the performance is increased more than two orders of magnitude compared to previously known micro-fabricated reactors for ammonia...... studies of both ammonia synthesis and decomposition, and it is shown how proper promotion can facilitate ammonia decomposition at temperatures below 500 K....

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

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.

Attenuation of thermal neutron through graphite

International Nuclear Information System (INIS)

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

2004-01-01

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

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

Photoemission study of K on graphite

NARCIS (Netherlands)

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

1999-01-01

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

Structural analysis of polycrystalline (graphitized) materials

International Nuclear Information System (INIS)

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

1993-01-01

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

Oxidation behavior of IG and NBG nuclear graphites

Energy Technology Data Exchange (ETDEWEB)

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

2011-01-15

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

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.

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

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

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

International Nuclear Information System (INIS)

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

1992-01-01

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

SWEPP PAN assay system uncertainty analysis: Passive mode measurements of graphite waste

International Nuclear Information System (INIS)

Blackwood, L.G.; Harker, Y.D.; Meachum, T.R.; Yoon, Woo Y.

1997-07-01

The Idaho National Engineering and Environmental Laboratory is being used as a temporary storage facility for transuranic waste generated by the U.S. Nuclear Weapons program at the Rocky Flats Plant (RFP) in Golden, Colorado. Currently, there is a large effort in progress to prepare to ship this waste to the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. In order to meet the TRU Waste Characterization Quality Assurance Program Plan nondestructive assay compliance requirements and quality assurance objectives, it is necessary to determine the total uncertainty of the radioassay results produced by the Stored Waste Examination Pilot Plant (SWEPP) Passive Active Neutron (PAN) radioassay system. To this end a modified statistical sampling and verification approach has been developed to determine the total uncertainty of a PAN measurement. In this approach the total performance of the PAN nondestructive assay system is simulated using computer models of the assay system and the resultant output is compared with the known input to assess the total uncertainty. This paper is one of a series of reports quantifying the results of the uncertainty analysis of the PAN system measurements for specific waste types and measurement modes. In particular this report covers passive mode measurements of weapons grade plutonium-contaminated graphite molds contained in 208 liter drums (waste code 300). The validity of the simulation approach is verified by comparing simulated output against results from measurements using known plutonium sources and a surrogate graphite waste form drum. For actual graphite waste form conditions, a set of 50 cases covering a statistical sampling of the conditions exhibited in graphite wastes was compiled using a Latin hypercube statistical sampling approach

Interrater and intrarater reliability of the Knosp scale for pituitary adenoma grading.

Science.gov (United States)

Mooney, Michael A; Hardesty, Douglas A; Sheehy, John P; Bird, Robert; Chapple, Kristina; White, William L; Little, Andrew S

2017-05-01

OBJECTIVE The goal of this study was to determine the interrater and intrarater reliability of the Knosp grading scale for predicting pituitary adenoma cavernous sinus (CS) involvement. METHODS Six independent raters (3 neurosurgery residents, 2 pituitary surgeons, and 1 neuroradiologist) participated in the study. Each rater scored 50 unique pituitary MRI scans (with contrast) of biopsy-proven pituitary adenoma. Reliabilities for the full scale were determined 3 ways: 1) using all 50 scans, 2) using scans with midrange scores versus end scores, and 3) using a dichotomized scale that reflects common clinical practice. The performance of resident raters was compared with that of faculty raters to assess the influence of training level on reliability. RESULTS Overall, the interrater reliability of the Knosp scale was "strong" (0.73, 95% CI 0.56-0.84). However, the percent agreement for all 6 reviewers was only 10% (26% for faculty members, 30% for residents). The reliability of the middle scores (i.e., average rated Knosp Grades 1 and 2) was "very weak" (0.18, 95% CI -0.27 to 0.56) and the percent agreement for all reviewers was only 5%. When the scale was dichotomized into tumors unlikely to have intraoperative CS involvement (Grades 0, 1, and 2) and those likely to have CS involvement (Grades 3 and 4), the reliability was "strong" (0.60, 95% CI 0.39-0.75) and the percent agreement for all raters improved to 60%. There was no significant difference in reliability between residents and faculty (residents 0.72, 95% CI 0.55-0.83 vs faculty 0.73, 95% CI 0.56-0.84). Intrarater reliability was moderate to strong and increased with the level of experience. CONCLUSIONS Although these findings suggest that the Knosp grading scale has acceptable interrater reliability overall, it raises important questions about the "very weak" reliability of the scale's middle grades. By dichotomizing the scale into clinically useful groups, the authors were able to address the poor

Reducing the Entrainment of Gangue Fines in Low Grade Microcrystalline Graphite Ore Flotation Using Multi-Stage Grinding-Flotation Process

Directory of Open Access Journals (Sweden)

Xiaoqing Weng

2017-03-01

Full Text Available A suitable grinding fineness and flow-sheet could potentially reduce the mechanical entrainment of gangue minerals in the flotation process of microcrystalline graphite. In this study, the suitable grinding fineness of a commercial graphite ore was estimated by mineralogy analysis and laboratory grind-flotation tests. The target grind size of this ore should be 92% passing 74 μm based on the mineralogical evaluation and the flotation performance. A comparison of a single-stage and a three-stage grinding circuit was conducted. Experimental results demonstrated that the three-stage grinding circuit could effectively improve the separation effect, which was attributed to the reduction of slimes. In the end, a more desirable beneficiation result was obtained with the application of three-stage grinding-flotation process by minimizing gangue entrainment.

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.

Mixed graphite cast iron for automotive exhaust component applications

Directory of Open Access Journals (Sweden)

De-lin Li

2017-11-01

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

Improving CS regulations.

Energy Technology Data Exchange (ETDEWEB)

Nesse, R.J.; Scheer, R.M.; Marasco, A.L.; Furey, R.

1980-10-01

President Carter issued Executive Order 12044 (3/28/78) that required all Federal agencies to distinguish between significant and insignificant regulations, and to determine whether a regulation will result in major impacts. This study gathered information on the impact of the order and the guidelines on the Office of Conservation and Solar Energy (CS) regulatory practices, investigated problems encountered by the CS staff when implementing the order and guidelines, and recommended solutions to resolve these problems. Major tasks accomplished and discussed are: (1) legislation, Executive Orders, and DOE Memoranda concerning Federal administrative procedures relevant to the development and analysis of regulations within CS reviewed; (2) relevant DOE Orders and Memoranda analyzed and key DOE and CS staff interviewed in order to accurately describe the current CS regulatory process; (3) DOE staff from the Office of the General Counsel, the Office of Policy and Evaluation, the Office of the Environment, and the Office of the Secretary interviewed to explore issues and problems encountered with current CS regulatory practices; (4) the regulatory processes at five other Federal agencies reviewed in order to see how other agencies have approached the regulatory process, dealt with specific regulatory problems, and responded to the Executive Order; and (5) based on the results of the preceding four tasks, recommendations for potential solutions to the CS regulatory problems developed. (MCW)

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)

Safety implications of a graphite oxidation accident in the compact ignition tokamak device

International Nuclear Information System (INIS)

Merrill, B.J.; O'Brien, M.H.

1989-01-01

This paper addresses the possible safety consequences of an air ingress accident for the Compact Ignition Tokamak (CIT) device. An experimental program was undertaken to determine oxidation rates of four nuclear grade graphites in air at temperatures ranging from 800 to 1800 C and flow velocities from 3 to 7 m/s. On the basis of these test results, an analytic model was developed to assess the extent of first wall/divertor protective tile oxidation and the amount of energy released from this oxidation. For CIT, a significant restriction to vacuum vessel air inflow will be provided by the air seals and walls of the surrounding test cells. Under these conditions, the graphite oxidation reaction inside the vacuum vessel will become oxygen starved within minutes of the onset of this event. Since significant oxidation rates were not achieved, the heat release did not elevate structural temperatures to levels of concern with regard to activated material release. 7 refs., 9 figs

Validation of CsNaIF data evaluation software. Final report on task FIN A940 on the Finnish support programme to IAEA safeguards

International Nuclear Information System (INIS)

Kaartinen, J.

1996-07-01

A new computer programme, called CsNaIF, which calculates the area of 137 Cs peak in spent fuel spectra has been developed for IAEA. This programme has been tested and evaluated in this report. Evaluation has been made by calculating different types of SFAT spectra (NaI- and CdTe-SFAT) with the validated software and with a research grade gamma spectroscopy software, SAMPO 90. Obtained results, mainly 137 Cs peak areas and their errors, have been compared and perceived differences have been reported. Also some recommendations of the usability of CsNaIF programme have been made for IAEA. (orig.) (4 refs.)

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.

NMR studies on graphite-methanol system

International Nuclear Information System (INIS)

El-Akkad, T.M.

1977-01-01

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

Mixed graphite cast iron for automotive exhaust component applications

OpenAIRE

De-lin Li

2017-01-01

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

Radiation damage in undoped CsI and CsI(Tl)

International Nuclear Information System (INIS)

Woody, C.L.; Kierstead, J.A.; Levy, P.W.; Stoll, S.

1992-01-01

Radiation damage has been studied in undoped CsI and CsI(TI) crystals using 60 Co gamma radiation for doses up to ∼ 4.2 x 10 6 . Samples from various manufacturers were measured ranging in size from 2.54 cm long cylinders to a 30 cm long block. Measurements were made on the change in optical transmission and scintillation light output as a function of dose. Although some samples showed a small change in transmission, a significant change in light output was observed for all samples. Recovery from damage was also studied as a function of time and exposure to UV light. A short lived phosphorescence was observed in undoped CsI, similar to the phosphorescence seen in CsI(TI)

Regeneration of Used Frying Palm Oil with Coffee Silverskin (CS), CS Ash (CSA) and Nanoparticles of CS (NCS).

Science.gov (United States)

Ismail, Samir Abd-Elmonem A; El-Anany, Ayman Mohammed; Ali, Rehab Farouk M

2017-01-01

The present investigation aimed to evaluate the efficiency of coffee silverskin (CS), CS ash (CSA) and nanoparticles of CS (NCS) in regeneration the quality of used frying palm oil. The adsorbents were mixed individually with used frying palm oil at level 4% (w/v) for 60 min. The properties of CS, CSA and NCS adsorbents were studied using (SEM) scanning electron microscopy technique. Some of physico-chemical characteristics of used frying palm oil (UFPO) and UFPO treated with adsorbents were determined. The results showed that the CS ash particles composed of irregular spherical and semispherical grains with deep cavities. The size of particles of CS ash ranged in diameter from 1.1 to 1.7 µm. The morphology of NCS consisted of cluster-type spherical nanoparticles and flakes. The particle size of NCS varies from 0.9 to 1.7 µm. Purification treatments caused marked (poil compared to untreated oil. The treatment of UFPO with 4% of adsorbents caused significant reductions in the content of free fatty acids ranged from 51.2 to 65.0%. The lowest level of peroxide (2.1 meq/kg) was recorded for UFPO treated with 4% of NCS. The highest reductions (72.8; 70.0%) in p-anisidine value were observed in UFPO treated with 4% of CSA and NCS, respectively. Treatment of UFPO with 4% of CS, CSA and NCS significantly lowered the polar content from 13.9% to 6.3, 4.8 and 3.9%, respectively. The results also indicate that CSA and NCS have nearly the same adsorption efficiency in lowering polymer content of UFPO. Filtration treatment of UFPO with 4% of CS, CSA and NCS markedly lowered the viscosity and colour values of treated UFPO.

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.

Raman characterization of bulk ferromagnetic nanostructured graphite

International Nuclear Information System (INIS)

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

2012-01-01

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

Fabrication of Graphene by Cleaving Graphite Chemically

Institute of Scientific and Technical Information of China (English)

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

2011-01-01

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

High temperature soldering of graphite

International Nuclear Information System (INIS)

Anikin, L.T.; Kravetskij, G.A.; Dergunova, V.S.

1977-01-01

The effect is studied of the brazing temperature on the strength of the brazed joint of graphite materials. In one case, iron and nickel are used as solder, and in another, molybdenum. The contact heating of the iron and nickel with the graphite has been studied in the temperature range of 1400-2400 ged C, and molybdenum, 2200-2600 deg C. The quality of the joints has been judged by the tensile strength at temperatures of 2500-2800 deg C and by the microstructure. An investigation into the kinetics of carbon dissolution in molten iron has shown that the failure of the graphite in contact with the iron melt is due to the incorporation of iron atoms in the interbase planes. The strength of a joint formed with the participation of the vapour-gas phase is 2.5 times higher than that of a joint obtained by graphite recrystallization through the carbon-containing metal melt. The critical temperatures are determined of graphite brazing with nickel, iron, and molybdenum interlayers, which sharply increase the strength of the brazed joint as a result of the formation of a vapour-gas phase and deposition of fine-crystal carbon

Properties of CsI and CsI-TMAE photocathodes

International Nuclear Information System (INIS)

Anderson, D.F.; Kwan, S.; Peskov, V.; Hoeneisen, B.

1992-06-01

The importance of heating the CsI or CsI-TMAE photocathodes during preparation, as well as the importance of the gas environment on the quantum efficiency is presented. The dependence of the aging characteristics of these photocathodes on the operating temperature, on the presence of gas, and on the charge amplification of the chamber is also discussed. For CsI photocathodes charges in excess of 2x10 14 e - /mm 2 can be collected with little degradation of performance. A timing resolution of 0.55 ns is also achieved for single photoelectrons suggesting a possible time-of-flight detector

Method of manufacturing a graphite coated fuel can

International Nuclear Information System (INIS)

Saito, Koichi; Uchida, Shunsuke.

1984-01-01

Purpose: To improve the close bondability and homogeneity of a graphite coating formed at the inner surface of a fuel can. Method: A coating containing graphite dispersed in a volatile organic solvent is used and a graphite coating is formed to the inner surface of a fuel can by way of a plunger method. After applying graphite coating, an inert gas is caused to flow at a certain flow rate to the inside of the fuel can horizontally rotaged so that gassification and evaporation of the volatile organic solvent contained in the graphite coating may be promoted. Since drying of the graphite coating coated to the inner surface of the fuel can thus be controlled, a graphite coating with satisfactory close bondability and homogeneity can be formed. (Kawakami, Y.)

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 .

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

Channel uranium-graphite reactor mounting

International Nuclear Information System (INIS)

Polushkin, K.K.; Kuznetsov, A.G.; Zheleznyakov, B.N.

1981-01-01

According to theoretical principles of general engineering technology the engineering experience of construction-mounting works at the NPP with channel uranium-graphite reactors is systematized. Main parameters and structural features of the 1000 MW channel uranium-graphite reactors are considered. The succession of mounting operations, premounting equipment and pipelines preparation and mounting works technique are described. The most efficient methods of fitting, welding and machining of reactor elements are recommended. Main problems of technical control service are discussed. A typical netted diagram of main equipment of channel uranium-graphite reactors mounting is given

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.

Concentration of 134Cs and 137Cs in Malaysian palm products

International Nuclear Information System (INIS)

Yii Mei Wo; Nor Aza Hassan; Narizan Sanusi

2008-08-01

Until today, countries importing food products from Malaysia, in region such as Indian Continental, Middle East, Central Europe and Central America still require the products to be tested for the radioactive contamination particularly for 134 Cs and 137 Cs. Information extracted from 3371 analytical results observed on isotopic concentration for 134 Cs and 137 Cs in the exported Malaysia origins palm products between years 2002 until 2007 had been summarized and reported in this work. Data shown that the concentration level of the concerned radio nuclides are below the minimum detectable activity and also well below maximum permissible limits adopted by Malaysian Government and also the international bodies. (Author)

Characterisation of Chlorine Behavior in French Graphite

International Nuclear Information System (INIS)

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

2016-01-01

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

Chemical sputtering of graphite by H+ ions

International Nuclear Information System (INIS)

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

1976-01-01

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

Separation of 134Cs and 137Cs from 125I solution for medical applications

International Nuclear Information System (INIS)

Ram, Ramu; Dash, Ashutosh; Banerjee, Dayamoy

2015-01-01

While neutron irradiation of natural Xe gas followed by wet chemical dissolution of activation products constitutes a successful paradigm for the small scale production 125 I, the concomitant production of 134 Cs and 137 Cs emerged as the primary impediment which necessitates purification of 125 I solution. This paper describes an ion-exchange chromatographic technique using Resorcinol Formaldehyde (RF) resin to purify 125 I solution from 134 Cs and 137 Cs impurities. A thorough investigation of the adsorption parameters of RF resin was carried out to arrive at the experimental conditions resulting optimum retention of 134 Cs and 137 Cs impurities. Based on the experimental findings, an optimized separation procedure was developed in which the neutron irradiated dissolved products at pH ∝ 13 was passed through a chromatography column containing RF resin where in 134 Cs and 137 Cs impurities gets adsorbed leaving behind 125 I to appear in the effluent. The overall recovery of 125 I was >90% with acceptable purity amenable for clinical applications.

A 2-D nucleation-growth model of spheroidal graphite

International Nuclear Information System (INIS)

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

2017-01-01

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

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.

Irradiation-induced amorphization process in graphite

Energy Technology Data Exchange (ETDEWEB)

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

1996-04-01

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

[Energy pooling collisions for K(4P) + Cs(5D) in a K-Cs mixture].

Science.gov (United States)

Aihemaiti, Pulati; Dai, Kang; Lu, Xin-hong; Shen, Yi-fan

2005-04-01

The rate coefficients for energy-pooling collisions K(4P) + Cs(5D) --> Cs(6S) + K(4D, 6S) in the K-Cs vapor mixture were measured relative to a known energy-pooling rate coefficient of a homonuclear reaction [i. e., Cs(6P) + Cs(5D) --> Cs(6S) + Cs (7D(J))]. Populations of the Cs(6P, 5D) and K(4P) states were produced by photodissociation of K2 and Cs2 molecules through the use of a dye laser radiation. The resulting fluorescence included the direct components emitted in the decay of the excited states produced by photodissociation and the induced components arising from the collisionally populated states. By combining relative intensities of the components with the effective lifetimes of Cs(6P) and K(4P) states, the rate coefficients (in units of 10(-9) cm3 x s(-1)) for the heteronuclear energy-pooling were found to be 2.6 and 3.6, respectively. The contribution to the rate coefficients from other processes are discussed.

Actinides in irradiated graphite of RBMK-1500 reactor

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-01

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

Structural features of the adsorption layer of pentacene on the graphite surface and the PMMA/graphite hybrid surface

Science.gov (United States)

Fadeeva, A. I.; Gorbunov, V. A.; Litunenko, T. A.

2017-08-01

Using the molecular dynamics and the Monte Carlo methods, we have studied the structural features and growth mechanism of the pentacene film on graphite and polymethylmethacrylate /graphite surfaces. Monolayer capacity and molecular area, optimal angles between the pentacene molecules and graphite and PMMA/graphite surfaces as well as the characteristic angles between the neighboring pentacene molecules in the adsorption layer were estimated. It is shown that the orientation of the pentacene molecules in the film is determined by a number of factors, including the surface concentration of the molecules, relief of the surface, presence or absence of the polymer layer and its thickness. The pentacene molecules adsorbed on the graphite surface keep a horizontal position relative to the long axis at any surface coverage/thickness of the film. In the presence of the PMMA layer on the graphite, the increase of the number of pentacene molecules as well as the thickness of the PMMA layer induce the change of molecular orientation from predominantly horizontal to vertical one. The reason for such behavior is supposed to be the roughness of the PMMA surface.

Changing US Attributes After CS-US Pairings Changes CS-Attribute-Assessments: Evidence for CS-US Associations in Attribute Conditioning.

Science.gov (United States)

Förderer, Sabine; Unkelbach, Christian

2016-03-01

Attribute Conditioning (AC) refers to people's changed assessments of stimuli's (CSs) attributes due to repeated pairing with stimuli (USs) possessing these attributes; for example, when an athletic person (US) is paired with a neutral person (CS), the neutral person is judged to be more athletic after the pairing. We hypothesize that this AC effect is due to CSs' associations with USs rather than direct associations with attributes. Three experiments test this hypothesis by changing US attributes after CS-US pairings. Experiments 1 and 2 conditioned athleticism by pairing neutral men (CSs) with athletic and non-athletic USs. Post-conditioning, USs' athleticism was reversed, which systematically influenced participants' assessment of CS athleticism. Experiment 3 conditioned athleticism and changed USs' musicality after CS-US pairings. This post-conditioning change affected musicality assessments of CSs but did not influence athleticism-assessments. The results indicate that AC effects are based on an associative CS-US-attribute structure. © 2016 by the Society for Personality and Social Psychology, Inc.

Changes in porosity of graphite caused by radiolytic gasification by carbon dioxide

International Nuclear Information System (INIS)

Murdie, Neil; Edwards, I.A.S.; Marsh, Harry

1986-01-01

Methods have been developed to study porosity in nuclear grade graphite. The changes induced during the radiolytic gasification of graphite in carbon dioxide have been investigated. Porosity in radiolytically gasified graphite (0-22.8% wt. loss) was examined by optical microscopy and scanning electron microscopy (SEM). Each sample was vacuum impregnated with a slow-setting resin containing a fluorescent dye. Optical microscopy was used to study pores >2 μm 2 c.s.a. A semi-automatic image analysis system linked to the optical microscope enabled pore parameter data including cross-sectional areas, perimeters, Feret's diameters and shape factors, to be collected. The results showed that radiolytic gasification produced a large increase in the number of pores 2 c.s.a. New open pores 2 c.s.a. were developed by gasification of existing open porosity into the closed porosity ( 2 c.s.a.) within the binder-coke. Open pores, 2-100 μm 2 c.s.a., which were gasified within the coarse-grained mosaics of the binder-coke. In the gasification process to 22.8% wt. loss, the apparent open pore volume increased from 6.6 to 33.8% and the apparent closed pore volumes decreased from approx. 3% to 0.1%. The increase in apparent open porosity from 6.6% (virgin) to 33.8% resulted from gasification within original open porosity and by the opening and development of closed porosity. There was no evidence for creation of porosity from within the 'bulk' graphite, it being developed from existing fine porosity. The structure of pores > 100 μm 2 c.s.a. showed no change because of the inhibition of oxidation by deposition of carbonaceous species from the CH 4 inhibitor. Such species diffuse to the pore wall and are sacrificially oxidised. (author)

A novel vanadosilicate with hexadeca-coordinated Cs+ ions as a highly effective Cs+ remover

International Nuclear Information System (INIS)

Datta, Shuvo Jit; Moon, Won Kyung; Choi, Do Young; Hwang, In Chul; Yoon, Kyung Byung

2014-01-01

The effective removal of 137 Cs + ions from contaminated groundwater and seawater and from radioactive nuclear waste solutions is crucial for public health and for the continuous operation of nuclear power plants. Various 137 Cs + removers have been developed, but more effective 137 Cs + removers are still needed. A novel microporous vanadosilicate with mixed-valence vanadium (V 4+ and V 5+ ) ions is now reported, which shows an excellent ability for Cs + capture and immobilization from groundwater, seawater, and nuclear waste solutions. This material is superior to other known materials in terms of selectivity, capacity, and kinetics, and at very low Cs + concentrations, it was found to be the most effective material for the removal of radioactive Cs + ions under the test conditions. This novel vanadosilicate also contains hexadeca-coordinated Cs + ions, which corresponds to the highest coordination number ever described.

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.

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)

Hybridization of Single Nanocrystals of Cs4PbBr6 and CsPbBr3.

Science.gov (United States)

Weerd, Chris de; Lin, Junhao; Gomez, Leyre; Fujiwara, Yasufumi; Suenaga, Kazutomo; Gregorkiewicz, Tom

2017-09-07

Nanocrystals of all-inorganic cesium lead halide perovskites (CsPbX 3 , X = Cl, Br, I) feature high absorption and efficient narrow-band emission which renders them promising for future generation of photovoltaic and optoelectronic devices. Colloidal ensembles of these nanocrystals can be conveniently prepared by chemical synthesis. However, in the case of CsPbBr 3 , its synthesis can also yield nanocrystals of Cs 4 PbBr 6 and the properties of the two are easily confused. Here, we investigate in detail the optical characteristics of simultaneously synthesized green-emitting CsPbBr 3 and insulating Cs 4 PbBr 6 nanocrystals. We demonstrate that, in this case, the two materials inevitably hybridize, forming nanoparticles with a spherical shape. The actual amount of these Cs 4 PbBr 6 nanocrystals and nanohybrids increases for synthesis at lower temperatures, i.e., the condition typically used for the development of perovskite CsPbBr 3 nanocrystals with smaller sizes. We use state-of-the-art electron energy loss spectroscopy to characterize nanoparticles at the single object level. This method allows distinguishing between optical characteristics of a pure Cs 4 PbBr 6 and CsPbBr 3 nanocrystal and their nanohybrid. In this way, we resolve some of the recent misconceptions concerning possible visible absorption and emission of Cs 4 PbBr 6 . Our method provides detailed structural characterization, and combined with modeling, we conclusively identify the nanospheres as CsPbBr 3 /Cs 4 PbBr 6 hybrids. We show that the two phases are independent of each other's presence and merge symbiotically. Herein, the optical characteristics of the parent materials are preserved, allowing for an increased absorption in the UV due to Cs 4 PbBr 6 , accompanied by the distinctive efficient green emission resulting from CsPbBr 3 .

Properties of Cs-intercalated single wall carbon nanotubes investigated by 133Cs Nuclear Magnetic resonance

KAUST Repository

Schmid, Marc R.

2012-11-01

In the present study, we investigated Cs-intercalated single wall carbon nanotubes (SWCNTs) using 133Cs Nuclear Magnetic resonance. We show that there are two types of Cs cations depending on the insertion level. Indeed, at low concentrations, Static spectra analysis shows that the Cs (α)+ species are fully ionized, i.e. α equal ca.1, while at higher concentrations a second paramagnetically shifted line appears, indicating the formation of Cs (β)+ ions with β < α ∼ +1. At low concentrations and low temperatures the Cs (α)+ ions exhibit a weak hyperfine coupling to the SWCNT conduction electrons, whereas, at higher temperatures, a thermally activated slow-motion diffusion process of the Cs (α)+ ions occurs along the interstitial channels present within the carbon nanotube bundles. At high concentrations, the Cs (β)+ ions seem to occupy well defined positions relative to the carbon lattice. As a matter of fact, the Korringa relaxation behavior suggests a strong hyperfine coupling between Cs nuclei and conduction electrons in the carbon nanotubes and a partial charge transfer, which suggest a plausible Cs(6s)-C(2p) hybridization. © 2012 Elsevier Ltd. All rights reserved.

Graphite oral tattoo: case report.

Science.gov (United States)

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

2015-10-16

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

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)

Fluorine determination in coal using high-resolution graphite furnace molecular absorption spectrometry and direct solid sample analysis

Energy Technology Data Exchange (ETDEWEB)

Machado, Patrícia M.; Morés, Silvane; Pereira, Éderson R. [Departamento de Química, Universidade Federal de Santa Catarina, 88040-900 Florianópolis, SC (Brazil); Welz, Bernhard, E-mail: w.bernardo@terra.com.br [Departamento de Química, Universidade Federal de Santa Catarina, 88040-900 Florianópolis, SC (Brazil); Instituto Nacional de Ciência e Tecnologia do CNPq, INCT de Energia e Ambiente, Universidade Federal da Bahia, 40170-115 Salvador, BA (Brazil); Carasek, Eduardo [Departamento de Química, Universidade Federal de Santa Catarina, 88040-900 Florianópolis, SC (Brazil); Andrade, Jailson B. de [Instituto Nacional de Ciência e Tecnologia do CNPq, INCT de Energia e Ambiente, Universidade Federal da Bahia, 40170-115 Salvador, BA (Brazil)

2015-03-01

The absorption of the calcium mono-fluoride (CaF) molecule has been employed in this study for the determination of fluorine in coal using direct solid sample analysis and high-resolution continuum source graphite furnace molecular absorption spectrometry (HR-CS GF MAS). The rotational line at 606.440 nm was used for measuring the molecular absorption in the gas phase. The pyrolysis and vaporization temperatures were 700 °C and 2100 °C, respectively. Different chemical modifiers have been studied, such as Pd and Ir as permanent modifiers, and Pd and the mixed Pd/Mg modifier in solution. The limit of detection and the characteristic mass were 0.3 and 0.1 ng F, respectively. One certified reference material (CRM) of coal (NIST 1635) and four CRMs with a non-certified value for F (SARM 18, SARM 20, BCR 40, BCR 180) were used to evaluate the accuracy and precision of the method, obtaining good agreement (104%) with the certified value and with the informed values (ranging from 90 to 103%). - Highlights: • High-resolution Graphite Furnace Molecular Absorption Spectrometry (HR-GF MAS) • Fluorine has been determined using HR-GF MAS of the CaF molecule. • The CaF molecule was generated in a graphite furnace at a temperature of 2100 °C • Coal samples have been analyzed using direct solid sample introduction. • Aqueous standard solutions have been used for calibration.

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)

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Design of the Graphite Reflectors in Research Reactors

Energy Technology Data Exchange (ETDEWEB)

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

2012-05-15

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

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.

Graphene-graphite oxide field-effect transistors.

Science.gov (United States)

Standley, Brian; Mendez, Anthony; Schmidgall, Emma; Bockrath, Marc

2012-03-14

Graphene's high mobility and two-dimensional nature make it an attractive material for field-effect transistors. Previous efforts in this area have used bulk gate dielectric materials such as SiO(2) or HfO(2). In contrast, we have studied the use of an ultrathin layered material, graphene's insulating analogue, graphite oxide. We have fabricated transistors comprising single or bilayer graphene channels, graphite oxide gate insulators, and metal top-gates. The graphite oxide layers show relatively minimal leakage at room temperature. The breakdown electric field of graphite oxide was found to be comparable to SiO(2), typically ~1-3 × 10(8) V/m, while its dielectric constant is slightly higher, κ ≈ 4.3. © 2012 American Chemical Society

Modification of structural graphite machining

International Nuclear Information System (INIS)

Lavrenev, M.M.

1979-01-01

Studied are machining procedures for structural graphites (GMZ, MG, MG-1, PPG) most widely used in industry, of the article mass being about 50 kg. Presented are dependences necessary for the calculation of cross sections of chip suction tappers and duster pipelines in machine shops for structural graphite machining

Friction anisotropy in boronated graphite

International Nuclear Information System (INIS)

Kumar, N.; Radhika, R.; Kozakov, A.T.; Pandian, R.; Chakravarty, S.; Ravindran, T.R.; Dash, S.; Tyagi, A.K.

2015-01-01

Graphical abstract: - Highlights: • Friction anisotropy in boronated graphite is observed in macroscopic sliding condition. • Low friction coefficient is observed in basal plane and becomes high in prismatic direction. • 3D phase of boronated graphite transformed into 2D structure after friction test. • Chemical activity is high in prismatic plane forming strong bonds between the sliding interfaces. - Abstract: Anisotropic friction behavior in macroscopic scale was observed in boronated graphite. Depending upon sliding speed and normal loads, this value was found to be in the range 0.1–0.35 in the direction of basal plane and becomes high 0.2–0.8 in prismatic face. Grazing-incidence X-ray diffraction analysis shows prominent reflection of (0 0 2) plane at basal and prismatic directions of boronated graphite. However, in both the wear tracks (1 1 0) plane become prominent and this transformation is induced by frictional energy. The structural transformation in wear tracks is supported by micro-Raman analysis which revealed that 3D phase of boronated graphite converted into a disordered 2D lattice structure. Thus, the structural aspect of disorder is similar in both the wear tracks and graphite transfer layers. Therefore, the crystallographic aspect is not adequate to explain anisotropic friction behavior. Results of X-ray photoelectron spectroscopy and Fourier transform infrared spectroscopy shows weak signature of oxygen complexes and functional groups in wear track of basal plane while these species dominate in prismatic direction. Abundance of these functional groups in prismatic plane indicates availability of chemically active sites tends to forming strong bonds between the sliding interfaces which eventually increases friction coefficient

Friction anisotropy in boronated graphite

Energy Technology Data Exchange (ETDEWEB)

Kumar, N., E-mail: niranjan@igcar.gov.in [Materials Science Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Radhika, R. [Crystal Growth Centre, Anna University, Chennai (India); Kozakov, A.T. [Research Institute of Physics, Southern Federal University, Rostov-on-Don (Russian Federation); Pandian, R. [Materials Science Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Chakravarty, S. [UGC-DAE CSR, Kalpakkam (India); Ravindran, T.R.; Dash, S.; Tyagi, A.K. [Materials Science Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

2015-01-01

Graphical abstract: - Highlights: • Friction anisotropy in boronated graphite is observed in macroscopic sliding condition. • Low friction coefficient is observed in basal plane and becomes high in prismatic direction. • 3D phase of boronated graphite transformed into 2D structure after friction test. • Chemical activity is high in prismatic plane forming strong bonds between the sliding interfaces. - Abstract: Anisotropic friction behavior in macroscopic scale was observed in boronated graphite. Depending upon sliding speed and normal loads, this value was found to be in the range 0.1–0.35 in the direction of basal plane and becomes high 0.2–0.8 in prismatic face. Grazing-incidence X-ray diffraction analysis shows prominent reflection of (0 0 2) plane at basal and prismatic directions of boronated graphite. However, in both the wear tracks (1 1 0) plane become prominent and this transformation is induced by frictional energy. The structural transformation in wear tracks is supported by micro-Raman analysis which revealed that 3D phase of boronated graphite converted into a disordered 2D lattice structure. Thus, the structural aspect of disorder is similar in both the wear tracks and graphite transfer layers. Therefore, the crystallographic aspect is not adequate to explain anisotropic friction behavior. Results of X-ray photoelectron spectroscopy and Fourier transform infrared spectroscopy shows weak signature of oxygen complexes and functional groups in wear track of basal plane while these species dominate in prismatic direction. Abundance of these functional groups in prismatic plane indicates availability of chemically active sites tends to forming strong bonds between the sliding interfaces which eventually increases friction coefficient.

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)

Structure and functionality of bromine doped graphite.

Science.gov (United States)

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

2013-04-28

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

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.

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...

Chemisputtering of interstellar graphite grains

International Nuclear Information System (INIS)

Draine, B.T.

1979-01-01

The rate of erosion of interstellar graphite grains as a result of chemical reaction with H, N, and O is estimated using the available experiment evidence. It is argued that ''chemical sputtering'' yields for interstellar graphite grains will be much less than unity, contrary to earlier estimates by Barlow and Silk. Chemical sputtering of graphite grains in evolving H II regions is found to be unimportant, except in extremely compact (n/sub H/> or approx. =10 5 cm -3 ) H II regions. Alternative explanations are considered for the apparent weakness of the lambda=2175 A extinction ''bump'' in the direction of several early type stars

Electronic properties of graphite

International Nuclear Information System (INIS)

Schneider, J.

2010-10-01

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

Low-energy electron observation of graphite and molybdenite crystals. Application to the study of graphite oxidation

International Nuclear Information System (INIS)

David, G.

1969-01-01

The LEED study of cleaved (0001) faces of crystals having a layered structure allowed to investigate flakes free of steps on graphite and molybdenite, to show twinning on natural graphite. By intensity measurements and computation in the case of a kinematical approximation it has been possible to determine an inner potential of 19 eV for graphite and to identify the direction of the Mo-S bond of the surface layer of molybdenite. The oxidation of graphite has been studied by observing changes, in symmetry of the diffraction patterns and by mass spectrometry of the gases evolved during the oxidation. No surface compounds have been detected and the carbon layers appeared to be peeled off one after the other. The oxidation took place at temperatures higher than 520 C under an oxygen pressure of 10 -5 torr. (author) [fr

Separation medium containing thermally exfoliated graphite oxide

Science.gov (United States)

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

2012-01-01

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

Influence of irradiation on high-strength graphites

International Nuclear Information System (INIS)

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

1989-01-01

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

Thermal gravimetric analysis of the CsCuCl3, Cs2CuCl4 and Cs2CuCl4x2H2O crystals

International Nuclear Information System (INIS)

Soboleva, L.V.; Vasil'eva, M.G.

1977-01-01

The thermal characteristics of crystals of Cs 2 CuCl 4 , Cs 2 CuCl 4 x2H 2 O, and CsCuCl 3 were investigated thermogravimetrically. The derivatogram of the Cs 2 CuCl 4 crystal is characterized by the presence of a single endothermal effect at 505 deg C. The derivatogram of the Cs 2 CuCl 4 x2H 2 O crystal contains three endothermal effects: at 40, 135, and 480 deg C. The derivatogram of the CsCuCl 3 crystal shows the presence of two endothermal effects at 142 and 455 deg C. The thermogravimetric data on Cs 2 CuCl 4 and CsCuCl 3 crystals reveal crystal decomposition on melting; hence, these crystals cannot be grown from melts

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

Use of Cs-137 for the calibration of the circulation model of Lithuanian coastal waters

International Nuclear Information System (INIS)

Davuliene, L.; Trinkunas, G.; Remeikis, V.; Valkunas, L.; Dick, S.

2002-01-01

It is well established that radioactive contamination of waters and sediments in the Lithuanian coastal area of the Baltic Sea is distributed unevenly. To describe the distribution of the radionuclides in waters of the Lithuanian coastal area of the Baltic Sea, the model based on the operational circulation model of the Bundesamt fuer Seeschffahrt und Hydrographie (BSH) for the North and Baltic seas was developed. The area under consideration contains both the Lithuanian coast of the Baltic Sea as well as the Curonian Gulf of the fresh water. The interplay between the salt and fresh water flows via the Kaipeda strait has impact on the distribution of radionuclides. For instance, Cs-137 is a typical radionuclide demonstrating this effect. It is experimentally well established that this radionuclide in the salt water is mainly in the dissolved form (about 90%) and just its minor part is concentrated in the suspended matter (about 10%). In fresh water the dissolved/suspended matter ratio for Cs-137 is totally opposite. Therefore, Cs-137 can be considered as the tracer following the fresh and salt water mixing. With samples of the radionuclide concentration in the sea area under consideration at hand, Cs-137 is used to normalize the tracer concentration simulated by the developed circulation model. The model was based on the grade of 1 nautic mile (nm), while the boundary conditions were taken from the more extended BSH model on the 6 nm grade. In order to understand the sensitivity of this local model to the initial conditions, the artificial conditions taken from the more general and coarse model were used. It has been obtained that the effect of the initial conditions is lost within 2-3 weeks. This result is independent of the coarse grain of the grade as calculations carried out on 1 nm and 0.5 nm grades show. The model was adopted for the PC Pentium III, and calculations of the salinity distribution depending on the meteorological conditions were carried out. Real

Collective modes in superconducting rhombohedral graphite

Energy Technology Data Exchange (ETDEWEB)

Kauppila, Ville [O.V. Lounasmaa Laboratory, Aalto University (Finland); Hyart, Timo; Heikkilae, Tero [University of Jyvaeskylae (Finland)

2015-07-01

Recently it was realized that coupling particles with a Dirac dispersion (such as electrons in graphene) can lead to a topologically protected state with flat band dispersion. Such a state could support superconductivity with unusually high critical temperatures. Perhaps the most promising way to realize such coupling in real materials is in the surface of rhombohedrally stacked graphite. We consider collective excitations (i.e. the Higgs modes) in surface superconducting rhombohedral graphite. We find two amplitude and two phase modes corresponding to the two surfaces of the graphite where the superconductivity lives. We calculate the dispersion of these modes. We also derive the Ginzburg-Landau theory for this material. We show that in superconducting rhombohedral graphite, the collective modes, unlike in conventional BCS superconductors, give a large contribution to thermodynamic properties of the material.

Computer Package for Graphite Total Cross-Section Calculations

International Nuclear Information System (INIS)

Adib, M.; Fathalla, M.

2008-01-01

An additive formula is given which allows calculating the contribution of the total neut.>neutron transmission through crystalline graphite. The formula takes into account the graphite form of poly or pyrolytic crystals and its parameters. Computer package Graphite has been designed in order to provide the required calculations in the neutron energy range from 0.1 MeV to 10 eV. The package includes three codes: PCG (Polycrystalline Graphite), PG (Pyrolytic Graphite) and HOPG (Highly Oriented Pyrolytic Graphite) for calculating neutron transmission through fine graphite powder (polycrystalline), neutron transmission and removal coefficient of PG crystal in terms of its mosaic spread for neutrons incident along its c-axis and the transmission of neutrons incident on HOPG crystal at different angles, respectively. For comparison of the experimental neutron transmission data with the calculated values, the program takes into consideration the effect of both wavelength and neutron beam divergence in either 2 constant wavelength spread mode (δλ=constant) or constant wavelength resolution mode (δλ/λ=constant). In order to check the validity for application of computer package Graphite in cross-section calculations, a comparison between calculated values with the available experimental data were carried out. An overall agreement is indicated with an accuracy sufficient for determine the neutron transmission characteristics

Design of graded filtration media in the diffusion-sedimentation regime

International Nuclear Information System (INIS)

Robinson, K.S.

1985-01-01

The collection of a graphite aerosol on silica gel particles has been studied experimentally. These data, and data from Gebhart et al., have been used to obtain an empirical correlation for penetration in terms of diffusion and sedimentation mechanisms and an interaction term. In addition, a rapid graphical procedure for the design of graded packed beds is presented based on the assumption that a linear axial deposition profile will maximize bed operating lifetime

Synthesis of carbon-13 labelled carbonaceous deposits and their evaluation for potential use as surrogates to better understand the behaviour of the carbon-14-containing deposit present in irradiated PGA graphite

Energy Technology Data Exchange (ETDEWEB)

Payne, L., E-mail: liam.payne@bristol.ac.uk [Interface Analysis Centre, HH Wills Physics Laboratory, University of Bristol, BS8 1TL (United Kingdom); Walker, S.; Bond, G. [Centre for Materials Science, University of Central Lancashire, PR1 2HE (United Kingdom); Eccles, H. [John Tyndall Institute for Nuclear Research, School of Computing, Engineering and Physical Sciences, University of Central Lancashire, PR1 2HE (United Kingdom); Heard, P.J.; Scott, T.B. [Interface Analysis Centre, HH Wills Physics Laboratory, University of Bristol, BS8 1TL (United Kingdom); Williams, S.J. [Radioactive Waste Management, B587, Curie Avenue, Harwell Oxford, Didcot, OX11 0RH (United Kingdom)

2016-03-15

The present work has used microwave plasma chemical vapour deposition to generate suitable isotopically labelled carbonaceous deposits on the surface of Pile Grade A graphite for use as surrogates for studying the behaviour of the deposits observed on irradiated graphite extracted from UK Magnox reactors. These deposits have been shown elsewhere to contain an enhanced concentration of {sup 14}C compared to the bulk graphite. A combination of Raman spectroscopy, ion beam milling with scanning electron microscopy and secondary ion mass spectrometry were used to determine topography and internal morphology in the formed deposits. Direct comparison was made against deposits found on irradiated graphite samples trepanned from a Magnox reactor core and showed a good similarity in appearance. This work suggests that the microwave plasma chemical vapour deposition technique is of value in producing simulant carbon deposits, being of sufficiently representative morphology for use in non-radioactive surrogate studies of post-disposal behaviour of {sup 14}C-containing deposits on some irradiated Magnox reactor graphite.

STS Observations of Landau Levels at Graphite Surfaces

OpenAIRE

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

2004-01-01

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

135Cs/137Cs isotopic composition of environmental samples across Europe: Environmental transport and source term emission applications

International Nuclear Information System (INIS)

Snow, Mathew S.; Snyder, Darin C.

2016-01-01

135 Cs/ 137 Cs isotopic analyses represent an important tool for studying the fate and transport of radiocesium in the environment; in this work the 135 Cs/ 137 Cs isotopic composition in environmental samples taken from across Europe is reported. Surface soil and vegetation samples from western Russia, Ukraine, Austria, and Hungary show consistent aged thermal fission product 135 Cs/ 137 Cs isotope ratios of 0.58 ± 0.01 (age corrected to 1/1/15), with the exception of one sample of soil-moss from Hungary which shows an elevated 135 Cs/ 137 Cs ratio of 1.78 ± 0.12. With the exception of the outlier sample from Hungary, surface soil/vegetation data are in quantitative agreement with values previously reported for soils within the Chernobyl exclusion zone, suggesting that radiocesium at these locations is primarily composed of homogenous airborne deposition from Chernobyl. Seawater samples taken from the Irish Sea show 135 Cs/ 137 Cs isotope ratios of 1.22 ± 0.11 (age corrected to 1/1/15), suggesting aged thermal fission product Cs discharged from Sellafield. The differences in 135 Cs/ 137 Cs isotope ratios between Sellafield, Chernobyl, and global nuclear weapons testing fallout indicate that 135 Cs/ 137 Cs isotope ratios can be utilized to discriminate between and track radiocesium transport from different nuclear production source terms, including major emission sources in Europe. - Highlights: • 135 Cs/ 137 Cs useful for tracking anthropogenic environmental radiocesium releases. • European surface soils/vegetation have uniform ratio consistent with Chernobyl. • 135 Cs/ 137 Cs in Irish sea represents thermal fission ratio distinct from Chernobyl. • Can distinguish between major source terms in Europe based on 135 Cs/ 137 Cs.

Reactive barriers for 137Cs retention

International Nuclear Information System (INIS)

Krumhansl, James L.; Brady, Patrick V.; Anderson, Howard L.

2000-01-01

137 Cs was dispersed globally by cold war activities and, more recently, by the Chernobyl accident. Engineered extraction of 137 Cs from soils and groundwaters is exceedingly difficult. Because the half life of 137 Cs is only 30.2 years, remediation might be more effective (and less costly) if 137 Cs bioavailability could be demonstrably limited for even a few decades by use of a reactive barrier. Essentially permanent isolation must be demonstrated in those few settings where high nuclear level wastes contaminated the environment with 135 Cs (half life 2.3x10 6 years) in addition to 137 Cs. Clays are potentially a low-cost barrier to Cs movement, though their long-term effectiveness remains untested. To identify optimal clays for Cs retention Cs resorption was measured for five common clays: Wyoming Montmorillonite (SWy-1), Georgia Kaolinites (KGa-1 and KGa-2), Fithian Illite (F-Ill), and K-Metabentonite (K-Mbt). Exchange sites were pre-saturated with 0.16 M CsCl for 14 days and readily exchangeable Cs was removed by a series of LiNO 3 and LiCl washes. Washed clay were then placed into dialysis bags and the Cs release to the deionized water outside the bags measured. Release rates from 75 to 139 days for SWy-1, K-Mbt and F- 111 were similar; 0.017 to 0.021% sorbed Cs released per day. Both kaolinites released Cs more rapidly (0.12 to 0.05% of the sorbed Cs per day). In a second set of experiments, clays were doped for 110 days and subjected to an extreme and prolonged rinsing process. All the clays exhibited some capacity for irreversible Cs uptake so most soils have some limited ability to act as a natural barrier to Cs migration. However, the residual loading was greatest on K-Mbt (∼ 0.33 wt% Cs). Thus, this clay would be the optimal material for constructing artificial reactive barriers

137Cs in man organism

International Nuclear Information System (INIS)

Marej, A.N.; Barkhudarov, R.M.

1980-01-01

Considered are the levels of 137 Cs content in the organism of adult urban population of the USSR and the main regularities of their alterations during 1962-1974. The non-uniform distribution of levels of 137 Cs buildup in adult population is shown to be connected with soil and geographic conditions. Food stuffs of local production have a noticeable effect on the levels of 137 Cs buildup in the organism of urban population. Calculation methods are used to study 137 Cs administration in a human organism. No significant difference is found between the indexes of 137 Cs content in the organism of rural and urban population. Presented are the levels of 137 Cs content in the organisms of inhabitants of such regions as Far North and Byelorussian-Ukrainian woodlands. Given are the reasons for the increased content of 137 Cs in the organism of rural population of these zones

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Superconductivity in graphite intercalation compounds

International Nuclear Information System (INIS)

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

2015-01-01

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

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.

Determination of Cl-36 in Irradiated Reactor Graphite

International Nuclear Information System (INIS)

Beer, H.-F.; Schumann, D.; Stowasser, T.; Hartmann, E.; Kramer, A.

2016-01-01

Two of the three research reactors at the Paul Scherrer Institute (PSI), the reactors DIORIT and PROTEUS, contained reactor graphite. Whereas the former research reactor DIORIT has been dismantled completely the PROTEUS is subject to a future decommissioning. In case of the DIORIT the reactor graphite was conditioned applying a procedure developed at PSI. In this case the 36 Cl content had to be determined after the conditioning. The result is reported in this paper. The radionuclide inventory including 36 Cl of the graphite used in PROTEUS was measured and the results are reported in here. It has been proven that the graphite from PROTEUS has a radionuclide inventory near the detection limits. All determined radionuclide activities are far below the Swiss exemptions limits. The graphite from PROTEUS therefore poses no radioactive waste. In contrast, the 36 Cl content of graphite from DIORIT is well above the exemption limits. (author)

Temperature distribution in graphite during annealing in air cooled reactors

International Nuclear Information System (INIS)

Oliveira Avila, C.R. de.

1989-01-01

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

Nanostructured carbon films with oriented graphitic planes

International Nuclear Information System (INIS)

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

2011-01-01

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

Principle design and data of graphite components

International Nuclear Information System (INIS)

Ishihara, Masahiro; Sumita, Junya; Shibata, Taiju; Iyoku, Tatsuo; Oku, Tatsuo

2004-01-01

The High Temperature Engineering Test Reactor (HTTR) constructed by Japan Atomic Energy Research Institute (JAERI) is a graphite-moderated and helium-gas-cooled reactor with prismatic fuel elements of hexagonal blocks. The reactor internal structures of the HTTR are mainly made up of graphite components. As well known, the graphite is a brittle material and there were no available design criteria for brittle materials. Therefore, JAERI had to develop the design criteria taking account of the brittle fracture behavior. In this paper, concept and key specification of the developed graphite design criteria is described, and also an outline of the quality control specified in the design criteria is mentioned

Technique for production of graphite-carbon products

Energy Technology Data Exchange (ETDEWEB)

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

1982-01-01

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

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.

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

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.)

Technical development of graphite waste treatment in NUPEC

International Nuclear Information System (INIS)

Saishu, S.; Inoue, T.

2001-01-01

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

Measurements of anomalous neutron transport in bulk graphite

Energy Technology Data Exchange (ETDEWEB)

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

2003-07-01

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

Exploring Polaronic, Excitonic Structures and Luminescence in Cs4PbBr6/CsPbBr3.

Science.gov (United States)

Kang, Byungkyun; Biswas, Koushik

2018-02-15

Among the important family of halide perovskites, one particular case of all-inorganic, 0-D Cs 4 PbBr 6 and 3-D CsPbBr 3 -based nanostructures and thin films is witnessing intense activity due to ultrafast luminescence with high quantum yield. To understand their emissive behavior, we use hybrid density functional calculations to first compare the ground-state electronic structure of the two prospective compounds. The dispersive band edges of CsPbBr 3 do not support self-trapped carriers, which agrees with reports of weak exciton binding energy and high photocurrent. The larger gap 0-D material Cs 4 PbBr 6 , however, reveals polaronic and excitonic features. We show that those lattice-coupled carriers are likely responsible for observed ultraviolet emission around ∼375 nm, reported in bulk Cs 4 PbBr 6 and Cs 4 PbBr 6 /CsPbBr 3 composites. Ionization potential calculations and estimates of type-I band alignment support the notion of quantum confinement leading to fast, green emission from CsPbBr 3 nanostructures embedded in Cs 4 PbBr 6 .

Nuclear graphite waste management. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

2001-05-01

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

Analysis of electrochemical disintegration process of graphite matrix

International Nuclear Information System (INIS)

Tian Lifang; Wen Mingfen; Chen Jing

2010-01-01

The electrochemical method with ammonium nitrate as electrolyte was studied to disintegrate the graphite matrix from the simulative fuel elements for high temperature gas-cooled reactor. The influences of process parameters, including salt concentration, system temperature and current density, on the disintegration rate of graphite fragments were investigated in the present work. The experimental results showed that the disintegration rate depended slightly on the temperature and salt concentration. The current density strongly affected the disintegration rate of graphite fragments. Furthermore, the content of introduced oxygen in final graphite fragments was independent of the current density and the concentration of electrolyte. Moreover, the structural evolution of graphite was analyzed based on the microstructural parameters determined by X-ray diffraction profile fitting analysis using MAUD (material analysis using diffraction) before and after the disintegration process. It may safely be concluded that the graphite disintegration can be ascribed to the influences of the intercalation of foreign molecules in between crystal planes and the partial oxidation involved. The disintegration process was described deeply composed of intercalate part and further oxidation part of carbon which effected together to lead to the collapse of graphite crystals.

Graphite moderator lifecycle behaviour. Proceedings of a specialists meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-08-01

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

Surface coating of graphite pebbles for Korean HCCR TBM

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

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

Surface coating of graphite pebbles for Korean HCCR TBM

International Nuclear Information System (INIS)

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

2014-01-01

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

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

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

Large-scale preparation of hollow graphitic carbon nanospheres

International Nuclear Information System (INIS)

Feng, Jun; Li, Fu; Bai, Yu-Jun; Han, Fu-Dong; Qi, Yong-Xin; Lun, Ning; Lu, Xi-Feng

2013-01-01

Hollow graphitic carbon nanospheres (HGCNSs) were synthesized on large scale by a simple reaction between glucose and Mg at 550 °C in an autoclave. Characterization by X-ray diffraction, Raman spectroscopy and transmission electron microscopy demonstrates the formation of HGCNSs with an average diameter of 10 nm or so and a wall thickness of a few graphenes. The HGCNSs exhibit a reversible capacity of 391 mAh g −1 after 60 cycles when used as anode materials for Li-ion batteries. -- Graphical abstract: Hollow graphitic carbon nanospheres could be prepared on large scale by the simple reaction between glucose and Mg at 550 °C, which exhibit superior electrochemical performance to graphite. Highlights: ► Hollow graphitic carbon nanospheres (HGCNSs) were prepared on large scale at 550 °C ► The preparation is simple, effective and eco-friendly. ► The in situ yielded MgO nanocrystals promote the graphitization. ► The HGCNSs exhibit superior electrochemical performance to graphite.

Calculated bond properties of K adsorbed on graphite

International Nuclear Information System (INIS)

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

1998-01-01

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

Low-energy electron observation of graphite and molybdenite crystals. Application to the study of graphite oxidation; Observation au moyen d'electrons de faible energie de cristaux de graphite et de molybdenite. Application a l'etude de l'oxydation du graphite

Energy Technology Data Exchange (ETDEWEB)

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

1969-07-01

The LEED study of cleaved (0001) faces of crystals having a layered structure allowed to investigate flakes free of steps on graphite and molybdenite, to show twinning on natural graphite. By intensity measurements and computation in the case of a kinematical approximation it has been possible to determine an inner potential of 19 eV for graphite and to identify the direction of the Mo-S bond of the surface layer of molybdenite. The oxidation of graphite has been studied by observing changes, in symmetry of the diffraction patterns and by mass spectrometry of the gases evolved during the oxidation. No surface compounds have been detected and the carbon layers appeared to be peeled off one after the other. The oxidation took place at temperatures higher than 520 C under an oxygen pressure of 10{sup -5} torr. (author) [French] L'etude par diffraction des electrons lents des faces (0001) de cristaux ayant une structure en feuillet a permis de mettre en evidence des plages sans gradins sur des clivages de graphite et de molybdenite caracterisees par la symetrie ternaire des diagrammes, de montrer l'existence de macles sur des cristaux de graphite naturel. Un calcul utilisant une approximation cinematique a ete applique aux intensites mesurees des taches de diffraction; il a ete ainsi possible de determiner un potentiel interne de 19 eV pour le graphite et de preciser la direction de la liaison Mo-S du feuillet superficiel de la molybdenite. L'oxydation du graphite a ete etudiee en mettant en relation des changements de symetrie des diagrammes de diffraction avec l'analyse des gaz provenant de la reaction carbone-oxygene. Il a ete montre qu'il n'y avait pas formation de composes de surface et que les couches de carbone etaient enlevees les unes apres les autres. L'oxydation a ete observee sous une pression d'oxygene de 10{sup -5} torr au-dessus de 520 C. (auteur)

Structural and Mössbauer spectroscopy characterization of bulk and nanostructured TiFe{sub 0.5} Ni{sub 0.5}/graphite compounds and their hydrides

Energy Technology Data Exchange (ETDEWEB)

Martínez, M. A. R., E-mail: fisicorodriguez@gmail.com; André-Filho, J.; Félix, L. L.; Coaquira, J. A. H.; Garg, V. K.; Oliveira, A. C. [Universidade de Brasília, Instituto de Física, Núcleo de Física Aplicada (Brazil); Mestnik-Filho, J. [Instituto de Pesquisas energéticas e Nucleares, IPEN-CNEN/SP (Brazil)

2015-06-15

The structural and hyperfine properties of bulk TiFe{sub 0.5}Ni{sub 0.5} intermetallic and ball-milled TiFe{sub 0.5}Ni{sub 0.5}/graphite compounds and their hydrides have been studied. The bulk and nanostructured TiFe{sub 0.5}Ni{sub 0.5} compounds crystallize in the cubic crystal structure of CsCl (B2). After hydrogenation, the formation of hydrogen-poor phase (∝-phase) and hydride phase (β-phase) have been determined for the bulk compound. However, the formation of the ∝-phase and the hydrogen-richest phase (γ-phase) and other secondary phases have been determined for the ball-milled TiFe{sub 0.5}Ni{sub 0.5}/graphite sample. It has been determined that the ball-milled TiFe{sub 0.5}Ni{sub 0.5}/graphite sample presents a large amount of the γ-phase which indicates that the presence of graphite nearby nanostructured intermetallic grains enhances the absorption of hydrogen. Mossbauer results are consistent with the structural results. Meanwhile, no significant changes in the isomer shift (IS) value has been determined for the α-phase with respect to the intermetallic compound, a strong increase in the IS value has been determined for the β- and γ-phases with respect to the ∝-phase. That increase indicates a decrease of the s-electron density at the Fe nuclei due to the charge transfer from the metal to the nearby hydrogen atoms.

Temperature and Pressure Dependence of the Reaction S plus CS (+M) -> CS2 (+M)

DEFF Research Database (Denmark)

Glarborg, Peter; Marshall, Paul; Troe, Juergen

2015-01-01

Experimental data for the unimolecular decomposition of CS2 from the literature are analyzed by unimolecular rate theory with the goal of obtaining rate constants for the reverse reaction S + CS (+M) -> CS2 (+M) over wide temperature and pressure ranges. The results constitute an important input...

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.

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)

Study on graphite samples for nuclear usage

International Nuclear Information System (INIS)

Suarez, J.C.M.; Silva Roseira, M. da

1994-01-01

Available as short communication only. The graphite, due to its properties (mechanical strength, thermal conductivity, high-temperature stability, machinability etc.) have many industrial applications, and consequently, an important strategic value. In the nuclear area, it has been used as moderator and reflector of neutrons in the fission process of uranium. The graphite can be produced from many types of carbonaceous materials by a variety of process dominated by the manufactures. This is the reason why there are in the world market a lot of graphite types with different physical and mechanical properties. The present investigation studies some physical characteristics of the graphite samples destined to use in a nuclear reactor. (author). 8 refs, 1 fig, 1 tab

Analysis of nuclear grade uranium oxides by atomic absorption spectrometry with electrothermal atomization

International Nuclear Information System (INIS)

Batistoni, D.A.; Erlijman, L.H.; Pazos, A.L.

1986-01-01

The application of atomic absorption spectrometry for the determination of five trace impurities in nuclear grade uranium oxides is described. The elements were separated from the uranium matrix by extraction chromatography and determined in 5.5 M nitric acid by electrothermal atomization using pyrolytic graphite coated tubes. Two elements, cadmium and chromium, with different volatility characteristics were employed to investigate the operating conditions. Drying and ashing conditions were studied for both elements. Ramp and constant potential (step) heating modes have also been studied and compared. Good reproducibility and a longer life of graphite tubes were obtained with ramp atomization. Detection limits (in micrograms per gram of uranium) were: Cd 0.01; Cr 0.1; Cu 0.4; Mn 0.04 and Ni 0.2. (author) [es

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.

A simple method for the deconvolution of 134 Cs/137 Cs peaks in gamma-ray scintillation spectrometry

International Nuclear Information System (INIS)

Darko, E.O.; Osae, E.K.; Schandorf, C.

1998-01-01

A simple method for the deconvolution of 134 Cs / 137 Cs peaks in a given mixture of 134 Cs and 137 Cs using Nal(TI) gamma-ray scintillation spectrometry is described. In this method the 795 keV energy of 134 Cs is used as a reference peak to calculate the activity of the 137 Cs directly from the measured peaks. Certified reference materials were measured using the method and compared with a high resolution gamma-ray spectrometry measurements. The results showed good agreement with the certified values. The method is very simple and does not need any complicated mathematics and computer programme to de- convolute the overlapping 604.7 keV and 661.6 keV peaks of 134 Cs and 137 Cs respectively. (author). 14 refs.; 1 tab., 2 figs

Intercomparison of graphite irradiations

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

While fast neutrons only are effective in damaging graphite, results of irradiations are more or less universally expressed in terms of thermal neutron fluxes. This paper attempts to correlate irradiations made in different reactors, i.e., in fluxes of different spectral compositions. Those attempts are based on comparison of 1) bulk length change and volume expansion, and 2) crystalline properties (e.g., lattice parameter C, magnetic susceptibility, stored energy, etc.). The methods used by various authors for determining the lattice constants of irradiated graphite are discussed. (author)

Relationship Between Active Learning Methodologies and Community College Students' STEM Course Grades

Science.gov (United States)

Clark Lesko, Cherish Christina

Active learning methodologies (ALM) are associated with student success, but little research on this topic has been pursued at the community college level. At a local community college, students in science, technology, engineering, and math (STEM) courses exhibited lower than average grades. The purpose of this study was to examine whether the use of ALM predicted STEM course grades while controlling for academic discipline, course level, and class size. The theoretical framework was Vygotsky's social constructivism. Descriptive statistics and multinomial logistic regression were performed on data collected through an anonymous survey of 74 instructors of 272 courses during the 2016 fall semester. Results indicated that students were more likely to achieve passing grades when instructors employed in-class, highly structured activities, and writing-based ALM, and were less likely to achieve passing grades when instructors employed project-based or online ALM. The odds ratios indicated strong positive effects (greater likelihoods of receiving As, Bs, or Cs in comparison to the grade of F) for writing-based ALM (39.1-43.3%, 95% CI [10.7-80.3%]), highly structured activities (16.4-22.2%, 95% CI [1.8-33.7%]), and in-class ALM (5.0-9.0%, 95% CI [0.6-13.8%]). Project-based and online ALM showed negative effects (lower likelihoods of receiving As, Bs, or Cs in comparison to the grade of F) with odds ratios of 15.7-20.9%, 95% CI [9.7-30.6%] and 16.1-20.4%, 95% CI [5.9-25.2%] respectively. A white paper was developed with recommendations for faculty development, computer skills assessment and training, and active research on writing-based ALM. Improving student grades and STEM course completion rates could lead to higher graduation rates and lower college costs for at-risk students by reducing course repetition and time to degree completion.

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.

Tunable CsPbBr3/Cs4PbBr6 phase transformation and their optical spectroscopic properties.

Science.gov (United States)

Chen, Xiao; Chen, Daqin; Li, Junni; Fang, Gaoliang; Sheng, Hongchao; Zhong, Jiasong

2018-04-24

As a novel type of promising materials, metal halide perovskites are a rising star in the field of optoelectronics. On this basis, a new frontier of zero-dimensional perovskite-related Cs4PbBr6 with bright green emission and high stability has attracted an enormous amount of attention, even though its photoluminescence still requires to clarification. Herein, the controllable phase transformation between three-dimensional CsPbBr3 and zero-dimensional Cs4PbBr6 is easily achieved in a facile ligand-assisted supersaturated recrystallization synthesis procedure via tuning the amount of surfactants, and their unique optical properties are investigated and compared in detail. Both Cs4PbBr6 and CsPbBr3 produce remarkably intense green luminescence with quantum yields up to 45% and 80%, respectively; however, significantly different emitting behaviors are observed. The fluorescence lifetime of Cs4PbBr6 is much longer than that of CsPbBr3, and photo-blinking is easily detected in the Cs4PbBr6 product, proving that the zero-dimensional Cs4PbBr6 is indeed a highly luminescent perovskite-related material. Additionally, for the first time, tunable emissions over the visible-light spectral region are demonstrated to be achievable via halogen composition modulations in the Cs4PbX6 (X = Cl, Br, I) samples. Our study brings a simple method for the phase control of CsPbBr3/Cs4PbBr6 and demonstrates the intrinsic luminescence nature of the zero-dimensional perovskite-related Cs4PbX6 products.

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

Graphite as negative electrode in Li-ion batteries; Le graphite comme electrode negative dans les accumulateurs Li-ion

Energy Technology Data Exchange (ETDEWEB)

Fischer, F.; Monnier, A. [Timcal SA (France)

1996-12-31

The last developments in lithium batteries design have demonstrated the advantages of graphite: competitive cost, flat output curve, high capacity thanks to the obtention of a final compound close to LiC{sub 6}, good behaviour during cycling and a high mass energy. However, these advantages are slightly tarnished by parasite secondary reactions during the evolution of the element. Two different cases are encountered: the formation of a passivation layer (loss of Li ions and formation of irreversible bounds) and the formation of a passivation layer with a reaction between graphite and the solvent (partial destruction of the graphite crystal lattice). In the first case, the theoretical graphite insertion capacity remains at 372 mAh/g while in the second case the insertion capacity is greatly reduced. Abstract only. (J.S.)

Graphite as negative electrode in Li-ion batteries; Le graphite comme electrode negative dans les accumulateurs Li-ion

Energy Technology Data Exchange (ETDEWEB)

Fischer, F; Monnier, A [Timcal SA (France)

1997-12-31

The last developments in lithium batteries design have demonstrated the advantages of graphite: competitive cost, flat output curve, high capacity thanks to the obtention of a final compound close to LiC{sub 6}, good behaviour during cycling and a high mass energy. However, these advantages are slightly tarnished by parasite secondary reactions during the evolution of the element. Two different cases are encountered: the formation of a passivation layer (loss of Li ions and formation of irreversible bounds) and the formation of a passivation layer with a reaction between graphite and the solvent (partial destruction of the graphite crystal lattice). In the first case, the theoretical graphite insertion capacity remains at 372 mAh/g while in the second case the insertion capacity is greatly reduced. Abstract only. (J.S.)

Microhydration of caesium compounds: Cs, CsOH, CsI and Cs₂I₂ complexes with one to three H₂O molecules of nuclear safety interest.

Science.gov (United States)

Sudolská, Mária; Cantrel, Laurent; Cernušák, Ivan

2014-04-01

Structure and thermodynamic properties (standard enthalpies of formation and Gibbs free energies) of hydrated caesium species of nuclear safety interest, Cs, CsOH, CsI and its dimer Cs₂I₂, with one up to three water molecules, are calculated to assess their possible existence in severe accident occurring to a pressurized water reactor. The calculations were performed using the coupled cluster theory including single, double and non-iterative triple substitutions (CCSD(T)) in conjunction with the basis sets (ANO-RCC) developed for scalar relativistic calculations. The second-order spin-free Douglas-Kroll-Hess Hamiltonian was used to account for the scalar relativistic effects. Thermodynamic properties obtained by these correlated ab initio calculations (entropies and thermal capacities at constant pressure as a function of temperature) are used in nuclear accident simulations using ASTEC/SOPHAEROS software. Interaction energies, standard enthalpies and Gibbs free energies of successive water molecules addition determine the ordering of the complexes. CsOH forms the most hydrated stable complexes followed by CsI, Cs₂I₂, and Cs. CsOH still exists in steam atmosphere even at quite high temperature, up to around 1100 K.

Graphite suspension in carbon dioxide

International Nuclear Information System (INIS)

Roche, R.

1965-01-01

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

1+ collective states of 124Cs and 126Cs nuclei

International Nuclear Information System (INIS)

Ivanova, S.P.; Kuliev, A.A.; Salamov, D.I.

1977-01-01

Within the framework of the random phase approximation β-decay properties of the 1 + states of 124 Cs and 126 Cs have been investigated. Greatly collectivized 1 + states in odd-odd nuclei are produced by the spin-dependent charge-exchange nucleon interaction. For numerical calculations the scheme of single-particle levels in the deformed Saxon-Woods potential has been used

Cesium platinide hydride 4Cs{sub 2}Pt.CsH: an intermetallic double salt featuring metal anions

Energy Technology Data Exchange (ETDEWEB)

Smetana, Volodymyr [Ames Laboratory, US Department of Energy, and Critical Materials Institute, Ames, Iowa, 50011-3020 (United States); Mudring, Anja-Verena [Ames Laboratory, US Department of Energy, and Critical Materials Institute, Ames, Iowa, 50011-3020 (United States); Department of Materials Sciences and Engineering, Iowa State University, Ames, Iowa, 50011-3111 (United States)

2016-11-14

With Cs{sub 9}Pt{sub 4}H a new representative of ionic compounds featuring metal anions can be added to this rare-membered family. Cs{sub 9}Pt{sub 4}H exhibits a complex crystal structure containing Cs{sup +} cations, Pt{sup 2-} and H{sup -} anions. Being a red, transparent compound its band gap is in the visible range of the electromagnetic spectrum and the ionic type of bonding is confirmed by quantum chemical calculations. This cesium platinide hydride can formally be considered as a double salt of the ''alloy'' cesium-platinum, or better cesium platinide, Cs{sub 2}Pt, and the salt cesium hydride CsH according to Cs{sub 9}Pt{sub 4}H≡4 Cs{sub 2}Pt.CsH. (copyright 2016 Wiley-VCH Verlag GmbH and Co. KGaA, Weinheim)

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)

Large-scale preparation of hollow graphitic carbon nanospheres

Energy Technology Data Exchange (ETDEWEB)

Feng, Jun; Li, Fu [Key Laboratory for Liquid-Solid Structural Evolution and Processing of Materials, Ministry of Education, Shandong University, Jinan 250061 (China); Bai, Yu-Jun, E-mail: byj97@126.com [Key Laboratory for Liquid-Solid Structural Evolution and Processing of Materials, Ministry of Education, Shandong University, Jinan 250061 (China); State Key laboratory of Crystal Materials, Shandong University, Jinan 250100 (China); Han, Fu-Dong; Qi, Yong-Xin; Lun, Ning [Key Laboratory for Liquid-Solid Structural Evolution and Processing of Materials, Ministry of Education, Shandong University, Jinan 250061 (China); Lu, Xi-Feng [Lunan Institute of Coal Chemical Engineering, Jining 272000 (China)

2013-01-15

Hollow graphitic carbon nanospheres (HGCNSs) were synthesized on large scale by a simple reaction between glucose and Mg at 550 Degree-Sign C in an autoclave. Characterization by X-ray diffraction, Raman spectroscopy and transmission electron microscopy demonstrates the formation of HGCNSs with an average diameter of 10 nm or so and a wall thickness of a few graphenes. The HGCNSs exhibit a reversible capacity of 391 mAh g{sup -1} after 60 cycles when used as anode materials for Li-ion batteries. -- Graphical abstract: Hollow graphitic carbon nanospheres could be prepared on large scale by the simple reaction between glucose and Mg at 550 Degree-Sign C, which exhibit superior electrochemical performance to graphite. Highlights: Black-Right-Pointing-Pointer Hollow graphitic carbon nanospheres (HGCNSs) were prepared on large scale at 550 Degree-Sign C Black-Right-Pointing-Pointer The preparation is simple, effective and eco-friendly. Black-Right-Pointing-Pointer The in situ yielded MgO nanocrystals promote the graphitization. Black-Right-Pointing-Pointer The HGCNSs exhibit superior electrochemical performance to graphite.

Knudsen cell--mass spectrometer studies of cesium--urania interactions. [Cs/sub 2/CO/sub 3/ or CsOH

Energy Technology Data Exchange (ETDEWEB)

Collins, J.L.; Osborne, M.F.; Malinauskas, A.P.; Lorenz, R.A.; Manning, S.R.

1976-06-01

Limited Knudsen cell--mass spectrometer studies were made of the partial pressures of cesium-containing species (assumed to be primarily Cs(g)) over Cs/sub 2/CO/sub 3/ and over phase equilibria involving UO/sub 2/ and probable Cs-U-O compounds formed from mixtures that initially contained either Cs/sub 2/CO/sub 3/-UO/sub 2/ or CsOH-UO/sub 2/. Although additional work is required to further define the equilibria involved, the data demonstrate unambiguously a significant reduction in cesium partial pressures due to probable Cs-U-O compound formation and indicate essentially identical behavior with either CsOH or Cs/sub 2/CO/sub 3/ as the starting material with UO/sub 2/.

Development of certified matrix reference materials for quality assurance of screening 134Cs and 137Cs in food

International Nuclear Information System (INIS)

Ishizu, H.; Yamada, T.

2013-01-01

A certified reference material using activated alumina powder certified for activity of 134 Cs and 137 Cs was developed. The results of the verification and the certification are described. The certified reference material can be used for quality assurance of screening activity measurements of 134 Cs and 137 Cs in food/foodstuffs. Commercially available equipments were experimentally tested using the CRM and another CRM including 40 K. The results of these tests are also shown. - Highlights: • CRM of 134 Cs and 137 Cs using activated alumina was developed. • CRM including 134 Cs, 137 Cs and 40 K was also developed. • Results of experimental performance test of commercial inspection equipments using CRMs were shown

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.

Method to Assess the Radionuclide Inventory of Irradiated Graphite from Gas-Cooled Reactors - 13072

Energy Technology Data Exchange (ETDEWEB)

Poncet, Bernard [EDF-CIDEN, 154 Avenue Thiers, CS 60018, F-69458 LYON cedex 06 (France)

2013-07-01

About 17,000 t of irradiated graphite waste will be produced from the decommissioning of the six French gas-cooled nuclear reactors. Determining the radionuclide (RN) content of this waste is of relevant importance for safety reasons and in order to determine the best way to manage them. For many reasons the impurity content that gave rise to the RNs in irradiated graphite by neutron activation during operation is not always well known and sometimes actually unknown. So, assessing the RN content by the use of traditional calculation activation, starting from assumed impurity content, leads to a false assessment. Moreover, radiochemical measurements exhibit very wide discrepancies especially on RN corresponding to precursor at the trace level such as natural chlorine corresponding to chlorine 36. This wide discrepancy is unavoidable and is due to very simple reasons. The level of impurity is very low because the uranium fuel used at that very moment was not enriched, so it was a necessity to have very pure nuclear grade graphite and the very low size of radiochemical sample is a simple technical constraint because device size used to get mineralization product for measurement purpose is limited. The assessment of a radionuclide inventory only based on few number of radiochemical measurements lead in most cases, to a gross over or under-estimation that is detrimental for graphite waste management. A method using an identification calculation-measurement process is proposed in order to assess a radiological inventory for disposal sizing purpose as precise as possible while guaranteeing its upper character. This method present a closer approach to the reality of the main phenomenon at the origin of RNs in a reactor, while also incorporating the secondary effects that can alter this result such as RN (or its precursor) release during reactor operation. (authors)

Crystallization degree change of expanded graphite by milling and annealing

International Nuclear Information System (INIS)

Tang Qunwei; Wu Jihuai; Sun Hui; Fang Shijun

2009-01-01

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

Ab Initio investigation of cesium monoxide of CsO and CsO+

International Nuclear Information System (INIS)

Zialenina, M.; Kelloe, V.; Cernusak, I.

2015-01-01

Cesium is material with a low work function and, accordingly, atomic Cs has a low value of ionization energy. Therefore cesium is regarded as a good source material for electrons in plasma heating module. One of plasma heating technologies using Cs grid is foreseen as a candidate for the tokamak within the framework of project ITER. Among the possible impurities that can coexist in this module are CsO or CsO + , due to presence of oxygen traces in the heating chamber. We conducted CCSD(T) energy calculations of the cesium oxide (X 2 Σ + ) and its cation (X 3 Σ - ). Here are presented the bond lengths and spectroscopic parameters of both species and ionization energy (IE). Our IE (6.88 eV) is in good agreement with previous theoretical results, experiment indicates substantially lower value (6.22 eV). (authors)

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)

Luminescence properties of undoped CsCaCl3 and CsSrCl3 crystalline scintillators

International Nuclear Information System (INIS)

Fujimoto, Yutaka; Saeki, Keiichiro; Koshimizu, Masanori; Asai, Keisuke; Yanagida, Takayuki

2015-01-01

Intrinsic luminescence properties of undoped CsCaCl 3 and CsSrCl 3 crystalline scintillators were studied. The crystal samples were grown by a vertical Bridgman method. Photoluminescence spectra of the crystals showed Auger-free luminescence (AFL) at 310 nm and self-trapped emission (STE) at 400 nm for CsCaCl 3 and 465 nm for CsSrCl 3 , when vacuum ultraviolet (VUV) light at 84 nm and 160 nm excited the crystals. X-ray excited radioluminescence spectra of the crystals showed some emission bands in the 280-600 nm wavelength range, which are owing to AFL, STE, and other origins such as lattice defects and impurities. Scintillation light yield was 400-300 ph/MeV, and the principal scintillation decay time about 2.5 ns and 12 ns for CsCaCl 3 and 1.8 ns and 13 ns for CsSrCl 3 . (author)

Adsorption behavior of bisphenol A on CTAB-modified graphite

Science.gov (United States)

Wang, Li-Cong; Ni, Xin-jiong; Cao, Yu-Hua; Cao, Guang-qun

2018-01-01

In this work, the adsorption behavior of BPA on CTAB-modified graphite was investigated thoroughly to develop a novel absorbent material. Atomic force microscopy revealed that conical admicelles formed on the surface of graphite. The surface area of graphite decreased significantly from 1.46 to 0.95 m2 g-1, which confirmed the formation of the larger size admicelle instead of the original smaller particle on the surface. CTAB concentration and incubation time affected the progress of admicelle formation on the surface of graphite. Adsolubilization is key in BPA adsorption by CTAB-modified graphite. An extraordinary cation-π electron interaction between CTAB and BPA, revealed by a red-shift in the ultraviolet spectrum, as well as a hydrophobic interaction contribute substantially to BPA adsolubilization. The equilibrium adsorption capacity of the modified graphite for BPA was 125.01 mg g-1. The adsorption kinetic curves of BPA on modified graphite were shown to follow a pseudosecond-order rate. The adsorption process was observed to be both spontaneous and exothermic complied with the Freundlich model.

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.

Fast CsI-phoswich detector

International Nuclear Information System (INIS)

Langenbrunner, J.R.

1996-01-01

An improved phoswich radiation detector used pure CsI crystal and a fast plastic scintillator and a single photomultiplier tube. The plastic is arranged to receive incident radiation, and that which passed through then strikes the CsI crystal. Scintillation light from both the plastic and CsI crystal are applied to the photomultiplier tube, with the light from the plastic passing through the crystal without absorption therein. Electronics are provided for analyzing the output of the photomultiplier tube to discriminate responses due to the plastic and the CsI crystal, through short gate and long gate integration, to produce results which are indicative of the characteristics of the different types of incident radiation, even in the presence of large amounts of radiation. The phoswich detector has excellent timing resolution. The scintillators of the CsI- phoswich were chosen for their fast risetimes, of about 3 ns for NE102A, and 30 ns for the pure CsI. 5 figs

Effect of Graphite on the Properties of Natural Rubber

Directory of Open Access Journals (Sweden)

Auda jabber Braihi

2016-09-01

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

Rapid determination of {sup 135}Cs and precise {sup 135}Cs/{sup 137}Cs atomic ratio in environmental samples by single-column chromatography coupled to triple-quadrupole inductively coupled plasma-mass spectrometry

Energy Technology Data Exchange (ETDEWEB)

Yang, Guosheng [Department of Radiation Chemistry, Institute of Radiation Emergency Medicine, Hirosaki University, 66-1 Hon-cho, Hirosaki, Aomori 036-8564 (Japan); Division of Nuclear Technology and Applications, Institute of High Energy Physics, Chinese Academy of Sciences, Beijing 100049 (China); Beijing Engineering Research Center of Radiographic Techniques and Equipment, Beijing 100049 (China); Tazoe, Hirofumi [Department of Radiation Chemistry, Institute of Radiation Emergency Medicine, Hirosaki University, 66-1 Hon-cho, Hirosaki, Aomori 036-8564 (Japan); Yamada, Masatoshi, E-mail: myamada@hirosaki-u.ac.jp [Department of Radiation Chemistry, Institute of Radiation Emergency Medicine, Hirosaki University, 66-1 Hon-cho, Hirosaki, Aomori 036-8564 (Japan)

2016-02-18

For source identification, measurement of {sup 135}Cs/{sup 137}Cs atomic ratio not only provides information apart from the detection of {sup 134}Cs and {sup 137}Cs, but it can also overcome the application limit that measurement of the {sup 134}Cs/{sup 137}Cs ratio has due to the short half-life of {sup 134}Cs (2.06 y). With the recent advancement of ICP-MS, it is necessary to improve the corresponding separation method for rapid and precise {sup 135}Cs/{sup 137}Cs atomic ratio analysis. A novel separation and purification technique was developed for the new generation of triple-quadrupole inductively coupled plasma-mass spectrometry (ICP-MS/MS). The simple chemical separation, incorporating ammonium molybdophosphate selective adsorption of Cs and subsequent single cation-exchange chromatography, removes the majority of isobaric and polyatomic interference elements. Subsequently, the ICP-MS/MS removes residual interference elements and eliminates the peak tailing effect of stable {sup 133}Cs, at m/z 134, 135, and 137. The developed analytical method was successfully applied to measure {sup 135}Cs/{sup 137}Cs atomic ratios and {sup 135}Cs activities in environmental samples (soil and sediment) for radiocesium source identification. - Highlights: • A simple {sup 135}Cs/{sup 137}Cs analytical method was developed. • The separation procedure was based on AMP adsorption and one column chromatography. • {sup 135}Cs/{sup 137}Cs was measured by ICP-MS/MS. • Decontamination factors for Ba, Mo, Sb, and Sn were improved. • {sup 135}Cs/{sup 137}Cs atomic ratios of 0.341–0.351 were found in Japanese soil samples.

From spent graphite to amorphous sp2+sp3 carbon-coated sp2 graphite for high-performance lithium ion batteries

Science.gov (United States)

Ma, Zhen; Zhuang, Yuchan; Deng, Yaoming; Song, Xiaona; Zuo, Xiaoxi; Xiao, Xin; Nan, Junmin

2018-02-01

Today, with the massive application of lithium ion batteries (LIBs) in the portable devices and electric vehicles, to supply the active materials with high-performances and then to recycle their wastes are two core issues for the development of LIBs. In this paper, the spent graphite (SG) in LIBs is used as raw materials to fabricate two comparative high-capacity graphite anode materials. Based on a microsurgery-like physical reconstruction, the reconstructed graphite (RG) with a sp2+sp3 carbon surface is prepared through a microwave exfoliation and subsequent spray drying process. In contrast, the neural-network-like amorphous sp2+sp3 carbon-coated graphite (AC@G) is synthesized using a self-reconfigurable chemical reaction strategy. Compared with SG and commercial graphite (CG), both RG and AC@G have enhanced specific capacities, from 311.2 mAh g-1 and 360.7 mAh g-1 to 409.7 mAh g-1 and 420.0 mAh g-1, at 0.1C after 100 cycles. In addition, they exhibit comparable cycling stability, rate capability, and voltage plateau with CG. Because the synthesis of RG and AC@G represents two typical physical and chemical methods for the recycling of SG, these results on the sp2+sp3 carbon layer coating bulk graphite also reveal an approach for the preparation of high-performance graphite anode materials derived from SG.

Adsorption of lead over graphite oxide.

Science.gov (United States)

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

2014-01-24

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

Interface structure between tetraglyme and graphite

Science.gov (United States)

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

2017-09-01

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

Dynamics of Supported Metal Nanoparticles Observed in a CS Corrected Environmental Transmission Electron Microscope

DEFF Research Database (Denmark)

Hansen, Thomas Willum; Dunin-Borkowski, Rafal E.; Wagner, Jakob Birkedal

resulting in the formation of larger particles and a loss of catalytic performance. Several models of sintering in different systems have been put forward [1,2]. However, most investigations have been post mortem studies, revealing only the final state of the catalyst. Transmission electron microscopy (TEM....... The combined capabilities of ETEM and image CS correction provide unique possibilities to study this relationship. However, in order to fully quantify image contrast from such experiments, a deeper understanding of the scattering of fast electrons in the presence of gas molecules in the pole piece gap...... of the microscope is needed. As industrial catalysts are usually complex high surface area materials, they are often not suited for fundamental studies. For this purpose, model systems consisting of gold nanoparticles on sheets of low surface area boron nitride and graphite supports were produced. Sheets...

Effect of reacting surface density on the overall graphite oxidation rate

International Nuclear Information System (INIS)

Oh, Chang; Kim, Eung; Lim, Jong; Schultz, Richard; Petti, David

2009-01-01

Graphite oxidation in an air-ingress accident is presently a very important issue for the reactor safety of the very high temperature gas cooled-reactor (VHTR), the concept of the next generation nuclear plant (NGNP) because of its potential problems such as mechanical degradation of the supporting graphite in the lower plenum of the VHTR might lead to core collapse if the countermeasure is taken carefully. The oxidation process of graphite has known to be affected by various factors, including temperature, pressure, oxygen concentration, types of graphite, graphite shape and size, flow distribution, etc. However, our recent study reveals that the internal pore characteristics play very important roles in the overall graphite oxidation rate. One of the main issues regarding graphite oxidation is the potential core collapse problem that may occur following the degradation of graphite mechanical strength. In analyzing this phenomenon, it is very important to understand the relationship between the degree of oxidization and strength degradation. In addition, the change of oxidation rate by graphite oxidation degree characterization by burn-off (ratio of the oxidized graphite density to the original density) should be quantified because graphite strength degradation is followed by graphite density decrease, which highly affects oxidation rates and patterns. Because the density change is proportional to the internal pore surface area, they should be quantified in advance. In order to understand the above issues, the following experiments were performed: (1) Experiment on the fracture of the oxidized graphite and validation of the previous correlations, (2) Experiment on the change of oxidation rate using graphite density and data collection, (3) Measure the BET surface area of the graphite. The experiments were performed using H451 (Great Lakes Carbon Corporation) and IG-110 (Toyo Tanso Co., Ltd) graphite. The reason for the use of those graphite materials is because

An alkali ion source based on graphite intercalation compounds for ion mobility spectrometry

International Nuclear Information System (INIS)

Tabrizchi, Mahmoud; Hosseini, Zahra S

2008-01-01

A variety of alkali cation emitters were developed as the ion source for ion mobility spectrometry. The cation emitters were constructed based on alkali ion graphite intercalation compounds (GICs). The compounds were prepared by fusing alkali salts with ground graphite. In order to produce alkali ions, the compounds were loaded on a filament and heated to red. Reactant ions of the form alk + ions were observed for the alkali salts NaCl, KCl.LiCl, CsCl and SrCl. In addition to Na + ions, K + ions were observed at the beginning of thermionic emission from Na-GIC. This is due to the low ionization potential of potassium that exists in trace amounts in sodium salts. In addition to the potassium ion, Na + was observed in the case of LiCl salt. The Na + and K + peaks originating from impurities totally disappeared after about 40 min. However, the thermionic emission of the main ion of the corresponding salt lasted for several days. No negative ions were observed upon reversing the drift field. Selected organic compounds (methyl isobutyl ketone, dimethyl sulfoxide, acetone and tetrahydrofuran) were also ionized via alkali cation attachment reaction. Distinct ion mobility patterns were observed for different substances using one type of alkali reactant ion. However, the ion mobility pattern for a given substance changed when a different alkali reactant ion was used. Ammonia and amines were not ionized when this source was used

A novel vanadosilicate with hexadeca-coordinated Cs{sup +} ions as a highly effective Cs{sup +} remover

Energy Technology Data Exchange (ETDEWEB)

Datta, Shuvo Jit; Moon, Won Kyung; Choi, Do Young; Hwang, In Chul; Yoon, Kyung Byung [Sogang Univ., Seoul (Korea, Republic of). Korea Center for Artificial Photosynthesis

2014-07-07

The effective removal of {sup 137}Cs{sup +} ions from contaminated groundwater and seawater and from radioactive nuclear waste solutions is crucial for public health and for the continuous operation of nuclear power plants. Various {sup 137}Cs{sup +} removers have been developed, but more effective {sup 137}Cs{sup +} removers are still needed. A novel microporous vanadosilicate with mixed-valence vanadium (V{sup 4+} and V{sup 5+}) ions is now reported, which shows an excellent ability for Cs{sup +} capture and immobilization from groundwater, seawater, and nuclear waste solutions. This material is superior to other known materials in terms of selectivity, capacity, and kinetics, and at very low Cs{sup +} concentrations, it was found to be the most effective material for the removal of radioactive Cs{sup +} ions under the test conditions. This novel vanadosilicate also contains hexadeca-coordinated Cs{sup +} ions, which corresponds to the highest coordination number ever described.

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.

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

A standard graphite block

Energy Technology Data Exchange (ETDEWEB)

Ivkovic, M; Zdravkovic, Z; Sotic, O [Department of Reactor Physics and Dynamics, Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1966-04-15

A graphite block was calibrated for the thermal neutron flux of the Ra-Be source using indium foils as detectors. Experimental values of the thermal neutron flux along the central vertical axis of the system were corrected for the self-shielding effect and depression of flux in the detector. The experimental values obtained were compared with the values calculated on the basis of solving the conservation neutron equation by the continuous slowing-down theory. In this theoretical calculation of the flux the Ra-Be source was divided into three resonance energy regions. The measurement of the thermal neutron diffusion length in the standard graphite block is described. The measurements were performed in the thermal neutron region of the system. The experimental results were interpreted by the diffusion theory for point thermal neutron source in the finite system. The thermal neutron diffusion length was calculated to be L= 50.9 {+-}3.1 cm for the following graphite characteristics: density = 1.7 g/cm{sup 3}; boron content = 0.1 ppm; absorption cross section = 3.7 mb.

A standard graphite block

International Nuclear Information System (INIS)

Ivkovic, M.; Zdravkovic, Z.; Sotic, O.

1966-04-01

A graphite block was calibrated for the thermal neutron flux of the Ra-Be source using indium foils as detectors. Experimental values of the thermal neutron flux along the central vertical axis of the system were corrected for the self-shielding effect and depression of flux in the detector. The experimental values obtained were compared with the values calculated on the basis of solving the conservation neutron equation by the continuous slowing-down theory. In this theoretical calculation of the flux the Ra-Be source was divided into three resonance energy regions. The measurement of the thermal neutron diffusion length in the standard graphite block is described. The measurements were performed in the thermal neutron region of the system. The experimental results were interpreted by the diffusion theory for point thermal neutron source in the finite system. The thermal neutron diffusion length was calculated to be L= 50.9 ±3.1 cm for the following graphite characteristics: density = 1.7 g/cm 3 ; boron content = 0.1 ppm; absorption cross section = 3.7 mb

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

Energy Technology Data Exchange (ETDEWEB)

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

1965-07-01

Since 1963 the Atomic Division of SNECMA has been conducting, under a contract with the CEA, an experimental work with a two-component fluid comprised of carbon dioxide and small graphite particles. The primary purpose was the determination of basic engineering information pertaining to the stability and the flowability of the suspension. The final form of the experimental loop consists mainly of the following items: a light-phase compressor, a heavy-phase pump, an electrical-resistance type heater section, a cooling heat exchanger, a hairpin loop, a transparent test section and a separator. During the course of the testing, it was observed that the fluid could be circulated quite easily in a broad range of variation of the suspension density and velocity - density from 30 to 170 kg/m{sup 3} and velocity from 2 to 24 m/s. The system could be restarted and circulation maintained without any difficulty, even with the heavy-phase pump alone. The graphite did not have a tendency to pack or agglomerate during operation. No graphite deposition was observed on the wall of the tubing. A long period run (250 hours) has shown the evolution of the particle dimensions. Starting with graphite of surface area around 20 m{sup 2}/g (graphite particles about 1 {mu}), the powder surface area reaches an asymptotic value of 300 m{sup 2}/g (all the particles less than 0.3 {mu}). Moisture effect on flow stability, flow distribution between two parallel channels, pressure drop in straight tubes, recompression ratio in diffusers were also investigated. (author) [French] Depuis 1963 la Division Atomique de la SNECMA conduit, dans le cadre d'un contrat avec le Commissariat A l'Energie Atomique, l'etude experimentale d'une suspension de fines particules de graphite dans le gaz carbonique. L'objectif principal est d'obtenir des informations d'ordre mecanique et technologique sur la mise en oeuvre de l'ecoulement de ce fluide diphase. Le circuit experimental comprend principalement: un

Determination of lead in rice grains by solid sampling HR-CS GFAAS.

Science.gov (United States)

Gunduz, Sema; Akman, Suleyman

2013-12-01

A study was performed for the determination of lead in rice grains directly by solid sampling high resolution continuum source graphite furnace atomic absorption spectrometry (HR-CS GFAAS) without digesting sample. The effects of sample amount, pyrolysis/atomization temperatures on the determination of lead in rice were investigated and optimized using a certificated rice flour. The lead concentrations were determined applying 800 °C of pyrolysis and 1800 °C of atomization temperatures without modifier. LOD (N=10, 3σ) and characteristic mass were 2.3 μg kg(-1) and 8.1 pg, respectively. The certified lead value of a rice flour CRM was found in its uncertainity limits. The lead contents of various rice samples obtained from markets in Turkey were between 0.009 and 0.162 mg kg(-1) which are in acceptable range. The average lead concentrations in various rice grains taken from the same package were significantly different from each other. Copyright © 2013 Elsevier Ltd. All rights reserved.

Pyrolysis and its potential use in nuclear graphite disposal

International Nuclear Information System (INIS)

Mason, J.B.; Bradbury, D.

2001-01-01

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

Tire containing thermally exfoliated graphite oxide

Science.gov (United States)

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

2011-01-01

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

Elaboration of aluminum oxide-based graphite containing castables

Science.gov (United States)

Zhou, Ningsheng

The aim of this work was set to develop effective and practicable new methods to incorporate natural flake graphite (FG) into the Al2O 3 based castables for iron and steel making applications. Three approaches, viz. micro-pelletized graphite (PG), crushed briquette of Al2O3-graphite (BAG) and TiO2 coated graphite (CFG), have been developed to insert flake graphite into Al2O 3 rich Al2O3-SiC based and Al2O 3-MgO based castables. These approaches were put into effect as countermeasures against the problems caused by FG in order: (1) to agglomerate the FG powders so as to decrease the specific surface area; (2) to diminish the density difference by using crushed carbon bonded compact of oxide-FG mixture; (3) to modify the surface of the flake graphite by forming hydrophilic coating; (4) to control the dispersion state of the graphite in the castable to maintain enough bonding strength; and (5) to use appropriate antioxidants to inhibit the oxidation of FG. The whole work was divided into two stages. In stage one, Al2O 3-SiC-C castables were dealt with to compare 4 modes of inserting graphite, i.e., by PG, BAG, CFG and FG. Overall properties were measured, all in correlation with graphite amount and incorporating mode. In stage two, efforts were made to reduce water demand in the Al2O3-MgO castables system. For this purpose, the matrix portion of the castable mixes was extracted and a coaxial double cylinder viscometer was adopted to investigate rheological characteristics of the matrix slurries vs. 4 kinds of deflocculants, through which the best deflocculant and its appropriate amount were found. Efforts were then made to add up to 30% MgO into the castables, using a limited amount of powders (antioxidants, Si, SiC, B4C and ZrB2, were added respectively or in combination. Overall properties of the castables, were investigated in correlation with MgO amount and graphite and antioxidant packages. Optimization work on oxidation and slag resistance was pursued. Finally

The behavior of radioactive 137Cs and stable Cs at the isolated undisturbed mountain pond in Fukui, Japan

International Nuclear Information System (INIS)

Iwamoto, Kazumi; Kimura, Makio; Ando, Kenji; Amano, Hikaru

2003-01-01

The behavior of radioactive 137 Cs and stable Cs at the isolated undisturbed mountain pond in Fukui, Japan was studied for the pond water, the sedimentary grains and the soil near the pond. The concentrations of 137 Cs and stable Cs in the pond water ranged from 0.23 to 0.85 Bq/m 3 and from 0.005 to 0.018 mg/m 3 , respectively. The sedimentary grains were sorted by sieving into fractions with diameter from 2 mm to less than 38 μm. The concentrations of 137 Cs and stable Cs in the sorted grains were measured, and those of the adsorbed state were determined by subtracting the concentration of the grain matrix. The adsorbed concentrations increased with decrease in particle diameter and depended less on the kind of samples. The in-situ distribution coefficient Kd depended largely on particle diameter and increased with the decrease in diameter. The values of Kd ranged from about 20 to 1200 m 3 /kg for stable Cs and about 15 to 1000 m 3 /kg for 137 Cs, and the Kd of 137 Cs seemed to be slightly smaller than that of stable Cs. The concentration of stable Cs in the sedimentary mud was found to be close to that of the fine grains. The concentrations of stable Cs in the soil near the pond was about 7.7 mg/kg, and that of 137 Cs was about 0.6 kBq/kg for the surface layer soil and decreased with increase in soil depth. (author)

The Godparent Plan: A Pedagogical Strategy for CS1 Accompaniment and CS2 Pedagogical Enhancement

Directory of Open Access Journals (Sweden)

Pedro Guillermo Feijóo-García

2018-02-01

Full Text Available Courses such as CS1 and CS2 can present an interesting pedagogical challenge when it comes to the theory-practice relationship, along with aspects that involve the course's logistics, the programming language used, and the characteristics of the students involved in the process. This study presents an innovative didactic approach, oriented towards the accompaniment of CS1 students by CS2 students at Universidad El Bosque, Colombia, seeking with this Godparent Plan, to provide a personalized accompaniment to first semester students, whereby CS2 students enhance their domain over concepts and skills while accompanying, explaining and teaching younger peers. The results of this study are favorable, outlining a didactic scheme that can be adapted and replicated in other curricular scenarios.

Late-time particle emission from laser-produced graphite plasma

Energy Technology Data Exchange (ETDEWEB)

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

2011-09-01

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

Late-time particle emission from laser-produced graphite plasma

International Nuclear Information System (INIS)

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

2011-01-01

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

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

Correlations between potassium, rubidium and cesium (133Cs and 137Cs) in sporocarps of Suillus variegatus in a Swedish boreal forest

International Nuclear Information System (INIS)

Vinichuk, M.; Rosen, K.; Johanson, K.J.; Dahlberg, A.

2011-01-01

An analysis of sporocarps of ectomycorrhizal fungi Suillus variegatus assessed whether cesium ( 133 Cs and 137 Cs) uptake was correlated with potassium (K) or rubidium (Rb) uptake. The question was whether intraspecific correlations of Rb, K and 133 Cs mass concentrations with 137 Cs activity concentrations in sporocarps were higher within, rather than among, different fungal species, and if genotypic origin of sporocarps within a population affected uptake and correlation. Sporocarps (n = 51) from a Swedish forest population affected by the fallout after the Chernobyl accident were studied. The concentrations were 31.9 ± 6.79 g kg -1 for K (mean ± SD, dwt), 0.40 ± 0.09 g kg -1 for Rb, 8.7 ± 4.36 mg kg -1 for 133 Cs and 63.7 ± 24.2 kBq kg -1 for 137 Cs. The mass concentrations of 133 Cs correlated with 137 Cs activity concentrations (r = 0.61). There was correlation between both 133 Cs concentrations (r = 0.75) and 137 Cs activity concentrations (r = 0.44) and Rb, but the 137 Cs/ 133 Cs isotopic ratio negatively correlated with Rb concentration. Concentrations of K and Rb were weakly correlated (r = 0.51). The 133 Cs mass concentrations, 137 Cs activity concentrations and 137 Cs/ 133 Cs isotopic ratios did not correlate with K concentrations. No differences between, within or, among genotypes in S. variegatus were found. This suggested the relationships between K, Rb, 133 Cs and 137 Cs in sporocarps of S. variegatus is similar to other fungal species. - Highlights: → We studied uptake of Cs ( 133 Cs and 137 Cs), K and Rb by Suillus variegates sporocarps. → Genotypic origin of fungus did not affect uptake of studied elements (isotopes). → Genotypic origin did not affect correlation between Cs ( 133 Cs and 137 Cs), K and Rb.

Preparation and Characterization of Graphite Waste/CeO2 Composites

Science.gov (United States)

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

2018-03-01

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

Radiation damage and life-time evaluation of RBMK graphite stack

Energy Technology Data Exchange (ETDEWEB)

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

1996-08-01

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

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

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.

Design Procedure of Graphite Components by ASME HTR Codes

International Nuclear Information System (INIS)

Kang, Ji-Ho; Jo, Chang Keun

2016-01-01

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

Measurement of thermal neutron cross section and resonance integral of the reaction {sup 135}Cs(n,{gamma}){sup 136}Cs

Energy Technology Data Exchange (ETDEWEB)

Katoh, Toshio; Nakamura, Shoji; Harada, Hideo [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan); Hatsukawa, Yuichi; Shinohara, Nobuo; Hata, Kentaro; Kobayashi, Katsutoshi; Motoishi, Shoji; Tanase, Masakazu

1997-03-01

The thermal neutron(2,200 m/s neutron) capture cross section({sigma}{sub 0}) and the resonance integral(I{sub 0}) of the reaction {sup 135}Cs(n,{gamma}){sup 136}Cs were measured by an activation method. Targets of radioactive cesium, which include {sup 135}Cs, {sup 137}Cs and stable {sup 133}Cs, were irradiated with reactor neutrons within or without a Cd shield case. The ratio of the number of nuclei of {sup 135}Cs to that of {sup 137}Cs was measured with a quadrupole mass spectrometer. This ratio and the ratio of activity of {sup 136}Cs to that of {sup 137}Cs were used for deduction of the {sigma}{sub 0} and the I{sub 0} of {sup 135}Cs. The {sigma}{sub 0} and the I{sub 0} of the reaction {sup 135}Cs(n,{sigma}){sup 136}Cs were 8.3 {+-} 0.3 barn and 38.1 {+-} 2.6 barn, respectively. (author)

Highly conductive bridges between graphite spheres to improve the cycle performance of a graphite anode in lithium-ion batteries

Energy Technology Data Exchange (ETDEWEB)

Wang, Hongyu [IM and T Ltd., Advanced Research Center, Saga University, Yoga-machi 1341, Saga 840-0047 (Japan); Umeno, Tatsuo; Mizuma, Koutarou [Research Center, Mitsui Mining Co. Ltd., Hibiki-machi 1-3, Wakamatsu-ku, Kitakyushu 808-0021 (Japan); Yoshio, Masaki [Advanced Research Center, Saga University, Yoga-machi 1341, Saga 840-0047 (Japan)

2008-01-10

Spherical carbon-coated natural graphite (SCCNG) is a promising anode material for lithium-ion batteries, but the smooth surface of graphite spheres is difficult to wet with an aqueous binder solution, and lacks electrical contacts. As a result, the cycle performance of such a graphite anode material is not satisfactory. An effective method has been introduced to tightly connect adjacent SCCNG particles by a highly conductive binder, viz. acetylene black bridges. The effect of the conductive bridges on the cyclability of SCCNG electrode has been investigated. (author)

Cs/CsPbX3 (X = Br, Cl) epitaxial heteronanocrystals with magic-angle stable/metastable grain boundary

Science.gov (United States)

Zhang, Yumeng; Fan, Baolu; Wu, Wenhui; Fan, Jiyang

2017-05-01

Metal-semiconductor heteronanostructures are crucial building blocks of nanoscale electronic and optoelectronic devices. However, the lattice misfit remains a challenge in constructing heteronanostructures. Perovskite nanocrystals are superior candidates for constructing nanodevices owing to excellent optical, ferroelectric, and superconducting properties. We report the epitaxial growth of lattice-matched Cs/CsPbBr3 metal-semiconductor heteronanocrystals in a liquid medium. The well-crystallized ultrathin Cs layers grow epitaxially on the surfaces of colloidal CsPbBr3 nanocrystals, forming heteronanocrystals with interface diameters of several nanometers. Most of them are pseudomorphic with coherent interfaces free from dislocations, and the others exhibit discrete high-angle grain boundaries. The model based on the calculation of the elastic potential energy of the epilayer and analysis of the near-coincidence sites explains well the experimental result. The analysis shows that the excellent lattice match between the metal and the semiconductor ensures the ideal epitaxial-growth of both Cs/CsPbBr3 and Cs/CsPbCl3 heteronanocrystals. Such metal/semiconductor heteronanocrystals pave the way for developing perovskite-based nanodevices.

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.)

Phase Transitions in CsSnCl3 and CsPbBr3 An NMR and NQR Study

Science.gov (United States)

Sharma, Surendra; Weiden, Norbert; Weiss, Alarich

1991-04-01

The phase transitions in CsSnCl3 and CsPbBr3 have been studied by X-ray powder diffraction, by 81Br-NQR and by 'H-, 119Sn-, and 113Cs-NMR. At room temperature in air CsSnCl3 forms a hydrate which can be dehydrated to the monoclinic phase II of CsSnCl3. The high temperature phase I has the Perovskite structure, as the X-ray and NMR experiments show. The three phases of CsPbBr3, known from literature, have been corroborated. The results are discussed in the framework of the group ABX3, A = alkalimetal ion, B = IV main group ion, and X = Halogen ion

VERA-CS Verification & Validation Plan

Energy Technology Data Exchange (ETDEWEB)

Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States)

2017-02-01

This report summarizes the current status of VERA-CS Verification and Validation for PWR Core Follow operation and proposes a multi-phase plan for continuing VERA-CS V&V in FY17 and FY18. The proposed plan recognizes the hierarchical nature of a multi-physics code system such as VERA-CS and the importance of first achieving an acceptable level of V&V on each of the single physics codes before focusing on the V&V of the coupled physics solution. The report summarizes the V&V of each of the single physics codes systems currently used for core follow analysis (ie MPACT, CTF, Multigroup Cross Section Generation, and BISON / Fuel Temperature Tables) and proposes specific actions to achieve a uniformly acceptable level of V&V in FY17. The report also recognizes the ongoing development of other codes important for PWR Core Follow (e.g. TIAMAT, MAMBA3D) and proposes Phase II (FY18) VERA-CS V&V activities in which those codes will also reach an acceptable level of V&V. The report then summarizes the current status of VERA-CS multi-physics V&V for PWR Core Follow and the ongoing PWR Core Follow V&V activities for FY17. An automated procedure and output data format is proposed for standardizing the output for core follow calculations and automatically generating tables and figures for the VERA-CS Latex file. A set of acceptance metrics is also proposed for the evaluation and assessment of core follow results that would be used within the script to automatically flag any results which require further analysis or more detailed explanation prior to being added to the VERA-CS validation base. After the Automation Scripts have been completed and tested using BEAVRS, the VERA-CS plan proposes the Watts Bar cycle depletion cases should be performed with the new cross section library and be included in the first draft of the new VERA-CS manual for release at the end of PoR15. Also, within the constraints imposed by the proprietary nature of plant data, as many as possible of the FY17

Epiphytic fruticose lichens as biomonitors for retrospective evaluation of the 134Cs/137Cs ratio in Fukushima fallout

International Nuclear Information System (INIS)

Ramzaev, V.; Barkovsky, A.; Gromov, A.; Ivanov, S.; Kaduka, M.

2014-01-01

In 2011–2013, sampling of epiphytic fruticose lichens of the genera Usnea, Bryoria and Alectoria was carried out on Sakhalin and Kuril Islands (the Sakhalin region, Russia) to investigate contamination of these organisms with the Fukushima-derived 134 Cs and 137 Cs. Activities of the radionuclides were determined in all 56 samples of lichens taken for the analysis. After correction for radioactive decay (on 15 March 2011), the activity concentrations ranged from 2.1 Bq kg −1 (d.w.) to 52 Bq kg −1 for 134 Cs and from 2.3 Bq kg −1 to 52 Bq kg −1 for 137 Cs. Cesium-134 and 137 Cs activities for the whole set of lichens (n = 56) were strongly positively correlated; Spearman's rank correlation coefficient was calculated as 0.991 (P < 0.01). The activity concentrations of 134 Cs and 137 Cs in Usnea lichens from the Sakhalin and Kunashir islands declined with a factor of three in the period from 2011 to 2013. The average biological half-time for both cesium radionuclides in lichens of the genus Usnea is estimated as 1.3 y. The mean of 0.99 ± 0.10 and median of 0.99 were calculated for the decay corrected 134 Cs/ 137 Cs activities ratios in the lichens (n = 56). The radionuclides ratio in the lichens did not depend on location of sampling site, species and the time that had passed after the Fukushima accident. The regression analysis has shown the background pre-Fukushima level of 137 Cs of 0.4 ± 0.3 Bq kg −1 , whereas the ratio between the Fukushima-borne 134 Cs and 137 Cs in the lichens was estimated as 1.04. The 134 Cs/ 137 Cs activities ratio in lichens from the Sakhalin region is consistent with the ratios reported by others for the heavy contaminated areas on Honshu Island in Japan following the Fukushima accident. The activity concentrations of natural 7 Be in lichens from the Sakhalin region varied between 100 Bq kg −1 and 600 Bq kg −1 ; the activity concentrations did not exhibit temporal variations during a 2y-period of observations. The

Methodology of characterization of radioactive graphite

International Nuclear Information System (INIS)

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

2014-01-01

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

Effect of waste mica on transfer factors of 134Cs to spinach and lettuce

International Nuclear Information System (INIS)

Sreenivasa Chari, M.; Manjaiah, K.M.; Sachdev, P.; Sachdev, M.S.

2011-01-01

A greenhouse pot culture experiment was conducted to study the effect of graded levels of waste mica (0, 10, 20 and 40 g kg -1 ) on reducing the radiocesium uptake by spinach (Spinacia olerecea L) and lettuce (Lactuca sativa L.) grown in 134 Cs-contaminated (at 37 k Bq kg -1 soil) Inceptisols, Vertisols and Ultisols. The biomass yield, and potassium content and its uptake by crops have been significantly improved by waste mica application. The crops grown in Vertisols recorded higher biomass yield, and K content and its uptake as compared with Inceptisols and Ultisols. The average 134 Cs transfer factor values recorded were: 0.21, 0.17 and 0.26 at the first cutting, 0.15, 0.12 and 0.28 at the second cutting and 0.07, 0.05 and 0.23 at the third cutting from Inceptisols, Vertisols and Ultisols, respectively. Waste mica significantly suppressed radiocesium uptake, the effect being more pronounced at 40 g mica kg -1 soil. There exists an inverse relationship between the 134 Cs transfer factors with plant potassium content and also the K uptake by the crops. (authors)