AGC-2 Graphite Pre-irradiation Data Package

DOE Office of Scientific and Technical Information (OSTI.GOV)

David Swank; Joseph Lord; David Rohrbaugh

2010-08-01

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 characterizedmore » 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.« less

Graphite for the nuclear industry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burchell, T.D.; Fuller, E.L.; Romanoski, G.R.

Graphite finds applications in both fission and fusion reactors. Fission reactors harness the energy liberated when heavy elements, such as uranium or plutonium, fragment or fission''. Reactors of this type have existed for nearly 50 years. The first nuclear fission reactor, Chicago Pile No. 1, was constructed of graphite under a football stand at Stagg Field, University of Chicago. Fusion energy devices will produce power by utilizing the energy produced when isotopes of the element hydrogen are fused together to form helium, the same reaction that powers our sun. The role of graphite is very different in these two reactormore » systems. Here we summarize the function of the graphite in fission and fusion reactors, detailing the reasons for their selection and discussing some of the challenges associated with their application in nuclear fission and fusion reactors. 10 refs., 15 figs., 1 tab.« less

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

NASA Astrophysics Data System (ADS)

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 impact is lower and can therefore be counteracted by temperature, a better reordering of the structure should be achieved. Concerning 14C, except when located close to open pores where it can be removed through radiolytic corrosion, it tends to stabilize in the graphite matrix into sp2 or sp3 structures with variable proportions depending on the irradiation conditions.

A new oxidation based technique for artifact free TEM specimen preparation of nuclear graphite

NASA Astrophysics Data System (ADS)

Johns, Steve; Shin, Wontak; Kane, Joshua J.; Windes, William E.; Ubic, Rick; Karthik, Chinnathambi

2018-07-01

Graphite is a key component in designs of current and future nuclear reactors whose in-service lifetimes are dependent upon the mechanical performance of the graphite. Irradiation damage from fast neutrons creates lattice defects which have a dynamic effect on the microstructure and mechanical properties of graphite. Transmission electron microscopy (TEM) can offer real-time monitoring of the dynamic atomic-level response of graphite subjected to irradiation; however, conventional TEM specimen-preparation techniques, such as argon ion milling itself, damage the graphite specimen and introduce lattice defects. It is impossible to distinguish these defects from the ones created by electron or neutron irradiation. To ensure that TEM specimens are artifact-free, a new oxidation-based technique has been developed. Bulk nuclear grades of graphite (IG-110 and NBG-18) and highly oriented pyrolytic graphite (HOPG) were initially mechanically thinned to ∼60 μm. Discs 3 mm in diameter were then oxidized at temperatures between 575 °C and 625 °C in oxidizing gasses using a new jet-polisher-like set-up in order to achieve optimal oxidation conditions to create self-supporting electron-transparent TEM specimens. The quality of these oxidized specimens were established using optical and electron microscopy. Samples oxidized at 575 °C exhibited large areas of electron transparency and the corresponding lattice imaging showed no apparent damage to the graphite lattice.

A new oxidation based technique for artifact free TEM specimen preparation of nuclear graphite

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johns, Steve; Shin, Wontak; Kane, Joshua J.

Graphite is a key component in designs of current and future nuclear reactors whose in-service lifetimes are dependent upon the mechanical performance of the graphite. Irradiation damage from fast neutrons creates lattice defects which have a dynamic effect on the microstructure and mechanical properties of graphite. Transmission electron microscopy (TEM) can offer real-time monitoring of the dynamic atomic-level response of graphite subjected to irradiation; however, conventional TEM specimen-preparation techniques, such as argon ion milling itself, damage the graphite specimen and introduce lattice defects. It is impossible to distinguish these defects from the ones created by electron or neutron irradiation. Thus,tomore » ensure that TEM specimens are artifact-free, a new oxidation-based technique has been developed. Bulk nuclear grades of graphite (IG-110 and NBG-18) and highly oriented pyrolytic graphite (HOPG) were initially mechanically thinned to ~60μm. Discs 3mm in diameter were then oxidized at temperatures between 575°C and 625°C in oxidizing gasses using a new jet-polisher-like set-up in order to achieve optimal oxidation conditions to create self-supporting electron-transparent TEM specimens. The quality of these oxidized specimens were established using optical and electron microscopy. Samples oxidized at 575°C exhibited large areas of electron transparency and the corresponding lattice imaging showed no apparent damage to the graphite lattice.« less

A new oxidation based technique for artifact free TEM specimen preparation of nuclear graphite

DOE PAGES

Johns, Steve; Shin, Wontak; Kane, Joshua J.; ...

2018-04-03

Graphite is a key component in designs of current and future nuclear reactors whose in-service lifetimes are dependent upon the mechanical performance of the graphite. Irradiation damage from fast neutrons creates lattice defects which have a dynamic effect on the microstructure and mechanical properties of graphite. Transmission electron microscopy (TEM) can offer real-time monitoring of the dynamic atomic-level response of graphite subjected to irradiation; however, conventional TEM specimen-preparation techniques, such as argon ion milling itself, damage the graphite specimen and introduce lattice defects. It is impossible to distinguish these defects from the ones created by electron or neutron irradiation. Thus,tomore » ensure that TEM specimens are artifact-free, a new oxidation-based technique has been developed. Bulk nuclear grades of graphite (IG-110 and NBG-18) and highly oriented pyrolytic graphite (HOPG) were initially mechanically thinned to ~60μm. Discs 3mm in diameter were then oxidized at temperatures between 575°C and 625°C in oxidizing gasses using a new jet-polisher-like set-up in order to achieve optimal oxidation conditions to create self-supporting electron-transparent TEM specimens. The quality of these oxidized specimens were established using optical and electron microscopy. Samples oxidized at 575°C exhibited large areas of electron transparency and the corresponding lattice imaging showed no apparent damage to the graphite lattice.« less

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

PubMed

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.

Damage tolerance of nuclear graphite at elevated temperatures

DOE PAGES

Liu, Dong; Gludovatz, Bernd; Barnard, Harold S.; ...

2017-06-30

Nuclear-grade graphite is a critically important high-temperature structural material for current and potentially next generation of fission reactors worldwide. It is imperative to understand its damage-tolerant behaviour and to discern the mechanisms of damage evolution under in-service conditions. Here we perform in situ mechanical testing with synchrotron X-ray computed micro-tomography at temperatures between ambient and 1,000 °C on a nuclear-grade Gilsocarbon graphite. We find that both the strength and fracture toughness of this graphite are improved at elevated temperature. Whereas this behaviour is consistent with observations of the closure of microcracks formed parallel to the covalent-sp 2-bonded graphene layers atmore » higher temperatures, which accommodate the more than tenfold larger thermal expansion perpendicular to these layers, we attribute the elevation in strength and toughness primarily to changes in the residual stress state at 800–1,000 °C, specifically to the reduction in significant levels of residual tensile stresses in the graphite that are ‘frozen-in’ following processing.« less

Damage tolerance of nuclear graphite at elevated temperatures

PubMed Central

Liu, Dong; Gludovatz, Bernd; Barnard, Harold S.; Kuball, Martin; Ritchie, Robert O.

2017-01-01

Nuclear-grade graphite is a critically important high-temperature structural material for current and potentially next generation of fission reactors worldwide. It is imperative to understand its damage-tolerant behaviour and to discern the mechanisms of damage evolution under in-service conditions. Here we perform in situ mechanical testing with synchrotron X-ray computed micro-tomography at temperatures between ambient and 1,000 °C on a nuclear-grade Gilsocarbon graphite. We find that both the strength and fracture toughness of this graphite are improved at elevated temperature. Whereas this behaviour is consistent with observations of the closure of microcracks formed parallel to the covalent-sp2-bonded graphene layers at higher temperatures, which accommodate the more than tenfold larger thermal expansion perpendicular to these layers, we attribute the elevation in strength and toughness primarily to changes in the residual stress state at 800–1,000 °C, specifically to the reduction in significant levels of residual tensile stresses in the graphite that are ‘frozen-in’ following processing. PMID:28665405

NEUTRONIC REACTOR

DOEpatents

Fermi, E.

1960-04-01

A nuclear reactor is described consisting of blocks of graphite arranged in layers, natural uranium bodies disposed in holes in alternate layers of graphite blocks, and coolant tubes disposed in the layers of graphite blocks which do not contain uranium.

Nondestructive evaluation of nuclear-grade graphite

NASA Astrophysics Data System (ADS)

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.

Effects of Oxidation on Oxidation-Resistant Graphite

DOE Office of Scientific and Technical Information (OSTI.GOV)

Windes, William; Smith, Rebecca; Carroll, Mark

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 oxidationmore » 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.« less

Understanding the reaction of nuclear graphite with molecular oxygen: Kinetics, transport, and structural evolution

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kane, Joshua J.; Contescu, Cristian I.; Smith, Rebecca E.

A thorough understanding of oxidation is important when considering the health and integrity of graphite components in graphite reactors. For the next generation of graphite reactors, HTGRs specifically, an unlikely air ingress has been deemed significant enough to have made its way into the licensing applications of many international licensing bodies. While a substantial body of literature exists on nuclear graphite oxidation in the presence of molecular oxygen and significant efforts have been made to characterize oxidation kinetics of various grades, the value of existing information is somewhat limited. Often, multiple competing processes, including reaction kinetics, mass transfer, and microstructuralmore » evolution, are lumped together into a single rate expression that limits the ability to translate this information to different conditions. This article reviews the reaction of graphite with molecular oxygen in terms of the reaction kinetics, gas transport, and microstructural evolution of graphite. It also presents the foundations of a model for the graphite-molecular oxygen reaction system that is kinetically independent of graphite grade, and is capable of describing both the bulk and local oxidation rates under a wide range of conditions applicable to air-ingress.« less

Understanding the reaction of nuclear graphite with molecular oxygen: Kinetics, transport, and structural evolution

DOE PAGES

Kane, Joshua J.; Contescu, Cristian I.; Smith, Rebecca E.; ...

2017-06-08

A thorough understanding of oxidation is important when considering the health and integrity of graphite components in graphite reactors. For the next generation of graphite reactors, HTGRs specifically, an unlikely air ingress has been deemed significant enough to have made its way into the licensing applications of many international licensing bodies. While a substantial body of literature exists on nuclear graphite oxidation in the presence of molecular oxygen and significant efforts have been made to characterize oxidation kinetics of various grades, the value of existing information is somewhat limited. Often, multiple competing processes, including reaction kinetics, mass transfer, and microstructuralmore » evolution, are lumped together into a single rate expression that limits the ability to translate this information to different conditions. This article reviews the reaction of graphite with molecular oxygen in terms of the reaction kinetics, gas transport, and microstructural evolution of graphite. It also presents the foundations of a model for the graphite-molecular oxygen reaction system that is kinetically independent of graphite grade, and is capable of describing both the bulk and local oxidation rates under a wide range of conditions applicable to air-ingress.« less

Characterization of nuclear graphite elastic properties using laser ultrasonic methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zeng, Fan W; Han, Karen; Olasov, Lauren R

2015-01-01

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

Measurements of print-through in graphite fiber epoxy composites

NASA Technical Reports Server (NTRS)

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

1989-01-01

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

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

NASA Astrophysics Data System (ADS)

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

2018-04-01

Graphite has been widely used as neutron moderator, reflector or fuel matrix in different types of reactors such as gas cooled nuclear reactors (UNGG, Magnox, AGR), RBMK reactors or high temperature gas cooled reactors. Their operation produces a great quantity of irradiated graphite or other carbonaceous waste (around 250,000 tons worldwide) that requires a special management strategy. In the case of disposal, which is a current management strategy, two main radionuclides, 14C and 36Cl might be dose determining at the outlet. Particular attention is paid to 14C due to its long half-life (T∼5730 years) [1] and as major contributor to the radioactive dose. 14C has two main production routes, i) transmutation of nitrogen (14N(n,p)14C) where nitrogen is mainly adsorbed at the surfaces of the irradiated graphite; ii) activation of carbon from the matrix (13C(n,γ)14C). According to leaching tests, it was shown that even if the quantity of 14C released in the solution is low (less than 1% of the initial inventory), around 30% is in the organic form that would be mobile in repository conditions [2,3]. 36Cl is mainly produced through the activation of 35Cl (35Cl(n,γ)36Cl) which is an impurity in nuclear graphite. Its activity is low but it might be highly mobile in clay host rocks. Thus, in order to make informed decisions about the best management process and to anticipate potential radionuclide dissemination during dismantling and in the repository, it is necessary to collect information on 14C and 36Cl location and speciation in graphite, after reactor closure. The goal of the present paper is therefore to use ion irradiation to simulate neutron irradiation and to evaluate the irradiation effects on the behavior of 36Cl and 14C as well as on the induced graphite structure modifications. For that, to understand and model the underlying mechanisms, we used an indirect approach based on 13C or 37Cl implantation to simulate the respective presence of 14C or 36Cl. These isotopes were implanted into Highly Oriented Pyrolytic Graphite (HOPG) samples used as a model material system representative of the nuclear graphite coke grains which form around 80% of nuclear graphite. Nuclear graphite is manufactured from petroleum coke grains (filler) blended with coal tar pitch acting as a binder. Shaped blocks are formed by extrusion of the blend. They are heat-treated up to about 2800 °C (graphitisation treatment) and polycrystalline graphite is obtained. Blocks, intended for the moderator or reflector, may be further impregnated with pitch, re-baked and regraphitised in order to increase the density. Virgin nuclear graphites have initial densities in the range 1.6-1.8 g cm-3. The difference with graphite crystal (density = 2.265 g cm-3) is due to internal porosity. As a result of mixing of several carbon compounds, this material is structurally heterogeneous at a local scale. Nuclear graphite presents a complex multiscale organisation. It can be locally more or less anisotropic and not completely graphitised. Nuclear graphite has a polycrystalline structure and contains micrometer sized grains. The grains are formed by several more or less oriented crystallites with a size of a few hundreds nanometers. Each crystallite is formed by a triperiodical stacking of graphene planes. Nuclear graphite contains also small amounts of impurities like oxygen, hydrogen, metals and halogens, among them chlorine [4]. Ion beam irradiation was used as a surrogate for neutrons because it may produce cascades (due to ballistic interactions) that could be similar to those created by neutrons in the nuclear reactor. Ion beam (or electron beam) irradiation has been used for many years to simulate neutron irradiation. It has advantages such as for example the possibility to vary the irradiation conditions and sometimes to carry out in situ observations. Moreover, depending on the ion nature and energy, it allows covering a broad range of the neutron recoil spectrum and the rate at which atoms are displaced can be increased in comparison to reactor conditions. Dose rates can thus be much higher than under neutron irradiation allowing for higher amounts of displacements per atoms (dpa) to be reached within some days instead of months or years. Moreover, because there is no sample activation, the samples are not radioactive [5-11]. During neutron irradiation, the neutrons interact with the matter both by collision with the atom nuclei (i.e. ballistic damage) and by nuclear reactions. The first atoms hit by neutrons are caused to move, thus starting a cascade of atomic collisions leading to electronic excitation as they go through the matter and on the path of the atoms they displace (recoil atoms). The ballistic damage can be evaluated using the nuclear stopping power and can be denoted by the number of displacements per atom (dpa). The effect of electronic excitation can be quantified using the electronic stopping power. The experimental simulation of neutron irradiation in a reactor can be done by irradiation of the graphite samples with different ions of different energies. The choice of these parameters enables the study of the damage effects with or without electron excitation or ballistic damage. Thus, knowing that the impinging neutrons induce mainly ballistic damage into the graphite matrix but that part of the recoil carbon energy is also transferred through electronic excitation, it is interesting to use ion irradiation because both ballistic damage and electronic excitation effects can be studied coupled or decoupled according to the nature of the ion, its energy and the fluence. It is possible to cover a wide range of electronic and nuclear stopping powers by working with different particle accelerators. Thus, we simulated the effects of these different irradiation regimes using ion irradiation by varying the Sn(nuclear)/Se(electronic) stopping power ratio as well as the irradiation temperature (from room temperature up to 1000 °C). Indeed, during reactor operation, neutron irradiation leads to changes in the graphite lattice parameters depending on irradiation conditions such as flux and fluence but also temperature [12]. Finally, Secondary Ion Mass Spectrometry (SIMS) analysis was used to determine 13C and 37Cl distribution profiles and allowed us to follow the implanted isotopes behavior. The structural modifications were followed by High Resolution Transmission Electron Microscopy (HRTEM) and Raman microspectrometry.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carroll, Mark C.

High-purity graphite is the core structural material of choice in the Very High Temperature Reactor (VHTR) design, a graphite-moderated, helium-cooled configuration capable of producing thermal energy for power generation as well as process heat for industrial applications that require temperatures higher than the outlet temperatures of present nuclear reactors. The Baseline Graphite Characterization Program is establishing accurate as-manufactured mechanical and physical property distributions in nuclear-grade graphites by providing comprehensive data that captures the level of variation in measured values. In addition to providing a thorough comparison between these values in different graphite grades, the program is also carefully tracking individualmore » specimen source, position, and orientation information in order to provide comparisons both in specific properties and in the associated variability between different lots, different billets, and different positions from within a single billet. This report is a preliminary comparison between each of the grades of graphite that are considered “candidate” grades from four major international graphite producers. These particular grades (NBG-18, NBG-17, PCEA, IG-110, and 2114) are the major focus of the evaluations presently underway on irradiated graphite properties through the series of Advanced Graphite Creep (AGC) experiments. NBG-18, a medium-grain pitch coke graphite from SGL from which billets are formed via vibration molding, was the favored structural material in the pebble-bed configuration. NBG-17 graphite from SGL is essentially NBG-18 with the grain size reduced by a factor of two. PCEA, petroleum coke graphite from GrafTech with a similar grain size to NBG-17, is formed via an extrusion process and was initially considered the favored grade for the prismatic layout. IG-110 and 2114, from Toyo Tanso and Mersen (formerly Carbone Lorraine), respectively, are fine-grain grades produced via an isomolding process. An analysis of the comparison between each of these grades will include not only the differences in fundamental and statistically-significant individual strength levels, but also the differences in the overall variability in properties within each of the grades that will ultimately provide the basis for predicting in-service performance. The comparative performance of the different types of nuclear-grade graphites will naturally continue to evolve as thousands more specimens are fully characterized with regard to strength, physical properties, and thermal performance from the numerous grades of graphite being evaluated.« less

Applications of X Radiography to Examination of Nuclear Graphite; APPLICATIONS DE LA RADIOGRAPHIE X A L'EXAMEN DU GRAPHITE NUCLEAIRE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Magnier, P.

1960-06-01

A technique which determines some important elements in the structure of graphite, osme dislocation lines, the presence of some dense impurities, and the local decreases in density, which develop in the course of oxidation, is described. (P.C.H.)

Neutronic reactor

DOEpatents

Carleton, John T.

1977-01-25

A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment.

Electron reflection and secondary emission characteristics of sputter-textured pyrolytic graphite surfaces

NASA Technical Reports Server (NTRS)

Wintucky, E. G.; Curren, A. N.; Sovey, J. S.

1981-01-01

Low secondary and reflected primary electron emission from the collector electrode surfaces is important for optimum collector efficiency and hence for high overall efficiency of microwave amplifier tubes used in communication satellites and in military systems. Ion sputter texturing of the surface effectively suppresses electron emission from pyrolytic graphite, which is a promising collector electrode material. Secondary and reflected primary electron emission characteristics of sputter textured pyrolytic graphite surfaces with microstructures of various sizes and densities are presented. The microstructure with the lowest electron emission levels, less than those of soot, consists of a dense array of tall, thin spires.

FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Davidson, J.K.

1963-11-19

A fuel element structure particularly useful in high temperature nuclear reactors is presented. Basically, the structure comprises two coaxial graphite sleeves integrally joined together by radial fins. Due to the high structural strength of graphite at high temperatures and the rigidity of this structure, nuclear fuel encased within the inner sleeve in contiguous relation therewith is supported and prevented from expanding radially at high temperatures. Thus, the necessity of relying on the usual cladding materials with relatively low temperature limitations for structural strength is removed. (AEC)

NUCLEAR REACTORS

DOEpatents

Long, E.; Ashley, J.W.

1958-12-16

A graphite moderator structure is described for a gas-cooled nuclear reactor having a vertical orlentation wherein the structure is physically stable with regard to dlmensional changes due to Wigner growth properties of the graphite, and leakage of coolant gas along spaces in the structure is reduced. The structure is comprised of stacks of unlform right prismatic graphite blocks positioned in layers extending in the direction of the lengths of the blocks, the adjacent end faces of the blocks being separated by pairs of tiles. The blocks and tiles have central bores which are in alignment when assembled and are provided with cooperatlng keys and keyways for physical stability.

6. VIEW OF INSIDE OF RAIL CAR CONTAINING GRAPHITE DELIVERED ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

6. VIEW OF INSIDE OF RAIL CAR CONTAINING GRAPHITE DELIVERED TO BUILDING 444. THE GRAPHITE WAS FORMED INTO MOLDS AND CRUCIBLE FOR USE IN THE FOUNDRY. (1/12/54) - Rocky Flats Plant, Non-Nuclear Production Facility, South of Cottonwood Avenue, west of Seventh Avenue & east of Building 460, Golden, Jefferson County, CO

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

NASA Astrophysics Data System (ADS)

Chi, Se-Hwan; Kim, Gen-Chan

2008-10-01

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

Neutronic reactor

DOEpatents

Lewis, Warren R.

1978-05-30

A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

Direct growth of nano-crystalline graphite films using pulsed laser deposition with in-situ monitoring based on reflection high-energy electron diffraction technique

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kwak, Jeong Hun; Lee, Sung Su; Lee, Hyeon Jun

2016-03-21

We report an experimental method to overcome the long processing time required for fabricating graphite films by a transfer process from a catalytic layer to a substrate, as well as our study of the growth process of graphite films using a pulsed laser deposition combined with in-situ monitoring based on reflection high-energy electron diffraction technique. We monitored the structural evolution of nano-crystalline graphite films directly grown on AlN-coated Si substrates without any catalytic layer. We found that the carbon films grown for less than 600 s cannot manifest the graphite structure due to a high defect density arising from grain boundaries;more » however, the carbon film can gradually become a nano-crystalline graphite film with a thickness of approximately up to 5 nm. The Raman spectra and electrical properties of carbon films indicate that the nano-crystalline graphite films can be fabricated, even at the growth temperature as low as 850 °C within 600 s.« less

Systems and methods for dismantling a nuclear reactor

DOEpatents

Heim, Robert R; Adams, Scott Ryan; Cole, Matthew Denver; Kirby, William E; Linnebur, Paul Damon

2014-10-28

Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.

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

NASA Astrophysics Data System (ADS)

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

2017-04-01

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

Nano-cracks in a synthetic graphite composite for nuclear applications

NASA Astrophysics Data System (ADS)

Liu, Dong; Cherns, David

2018-05-01

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

The Use of Basalt, Basalt Fibers and Modified Graphite for Nuclear Waste Repository - 12150

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gulik, V.I.; Biland, A.B.

2012-07-01

New materials enhancing the isolation of radioactive waste and spent nuclear fuel are continuously being developed.. Our research suggests that basalt-based materials, including basalt roving chopped basalt fiber strands, basalt composite rebar and materials based on modified graphite, could be used for enhancing radioactive waste isolation during the storage and disposal phases and maintaining it during a significant portion of the post-closure phase. The basalt vitrification process of nuclear waste is a viable alternative to glass vitrification. Basalt roving, chopped basalt fiber strands and basalt composite rebars can significantly increase the strength and safety characteristics of nuclear waste and spentmore » nuclear fuel storages. Materials based on MG are optimal waterproofing materials for nuclear waste containers. (authors)« less

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

NASA Astrophysics Data System (ADS)

Nyathi, Mhlwazi S.

2011-12-01

Graphite is utilized as a neutron moderator and structural component in some nuclear reactor designs. During the reactor operaction the structure of graphite is damaged by collision with fast neutrons. Graphite's resistance to this damage determines its lifetime in the reactor. On neutron irradiation, isotropic or near-isotropic graphite experiences less structural damage than anisotropic graphite. The degree of anisotropy in a graphite artifact is dependent on the structure of its precursor coke. Currently, there exist concerns over a short supply of traditional precursor coke, primarily due to a steadily increasing price of petroleum. The main goal of this study was to study the anisotropic and isotropic properties of graphitized co-cokes and anthracites as a way of investigating the possibility of synthesizing isotropic or near-isotropic graphite from co-cokes and anthracites. Demonstrating the ability to form isotropic or near-isotropic graphite would mean that co-cokes and anthracites have a potential use as filler material in the synthesis of nuclear graphite. The approach used to control the co-coke structure was to vary the reaction conditions. Co-cokes were produced by coking 4:1 blends of vacuum resid/coal and decant oil/coal at temperatures of 465 and 500 °C for reaction times of 12 and 18 hours under autogenous pressure. Co-cokes obtained were calcined at 1420 °C and graphitized at 3000 °C for 24 hours. Optical microscopy, X-ray diffraction, temperature-programmed oxidation and Raman spectroscopy were used to characterize the products. It was found that higher reaction temperature (500 °C) or shorter reaction time (12 hours) leads to an increase in co-coke structural disorder and an increase in the amount of mosaic carbon at the expense of textural components that are necessary for the formation of anisotropic structure, namely, domains and flow domains. Characterization of graphitized co-cokes showed that the quality, as expressed by the degree of graphitization and crystallite dimensions, of the final product is dependent on the nature of the precursor co-coke. The methodology for studying anthracites was to select two anthracites on basis of rank, PSOC1515 being semi-anthracite and DECS21 anthracite. The selected anthracites were graphitized, in both native and demineralized states, under the same conditions as co-cokes. Products obtained from DECS21 showed higher degrees of graphitization and larger crystallite dimensions than products obtained from PSOC1515. Demineralization of anthracites served to increase the degree of graphitization, indicating that the minerals contained in these anthracites have no graphitization-enhancing ability. A larger crystallite length for products obtained from native versions, compared to demineralized versions, was attributed to a formation and decomposition of a silicon carbide during graphitization of native versions. In order to examine the anisotropic and isotropic properties, nuclear-grade graphite samples obtained from Oak Ridge National Laboratory (ORNL) and commercial graphite purchased from Fluka were characterized under similar conditions as graphitized co-cokes and anthracites. These samples served as representatives of "two extremes", with ORNL samples being the isotropic end and commercial graphite being the anisotropic end. Through evaluating relationships between structural parameters, it was observed that graphitized co-cokes are situated, structurally, somewhere between the "two extremes", whereas graphitized anthracites are closer to the anisotropic end. Basically, co-cokes have a better potential than anthracites to transform to isotropic or near-isotropic graphite upon graphitization. By co-coking vacuum resid/coal instead of decant oil/coal or using 500 °C instead of 465 °C, a shift away from commercial graphite towards ORNL samples was attained. Graphitizing a semi-anthracite or demineralizing anthracites before graphitization also caused a shift towards ORNL samples.

NUCLEAR REACTORS

DOEpatents

Long, E.; Ashby, J.W.

1958-09-16

ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

Effect of Reacting Surface Density on the Overall Graphite Oxidation Rate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chang H. Oh; Eung Kim; Jong Lim

2009-05-01

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 internalmore » 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 their chemical and mechanical characteristics are well identified by the previous investigations, and therefore it was convenient for us to access the published data, and to apply and validate our new methodologies. This paper presents preliminary results of compressive strength vs. burn-off and surface area density vs. burn-off, which can be used for the nuclear graphite selection for the NGNP.« less

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

PubMed

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

2017-05-01

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

Thermophysical property and pore structure evolution in stressed and non-stressed neutron irradiated IG-110 nuclear graphite

DOE Office of Scientific and Technical Information (OSTI.GOV)

Snead, Lance; Contescu, Christian I.; Byun, Thak Sang

2016-08-01

The nuclear graphite, IG-110, was irradiated with and without a compressive load of 5 MPa at ~400 *C up to 9.3E25 n/m2 (E > 0.1 MeV). Following irradiation physical properties were studied to compare the effect of graphite irradiation on microstructure developed under compression and in stress-free conditions. Properties included: dimensional change, thermal conductivity, dynamic modulus, and CTE. The effect of stress on open internal porosity was determined through nitrogen adsorption. The IG-110 graphite experienced irradiation-induced creep that is differentiated from irradiation-induced swelling. Irradiation under stress resulted in somewhat greater thermal conductivity and coefficient of thermal expansion. While a significantmore » increase in dynamic modulus occurs, no differentiation between materials irradiated with and without compressive stress was observed. Nitrogen adsorption analysis suggests a difference in pore evolution in the 0.3e40 nm range for graphite irradiated with and without stress, but this evolution is seen to be a small contributor to the overall dimensional change.« less

Thermophysical property and pore structure evolution in stressed and non-stressed neutron irradiated IG-110 nuclear graphite

DOE PAGES

Snead, Lance L.; Contescu, C. I.; Byun, T. S.; ...

2016-04-23

The nuclear graphite, IG-110, was irradiated with and without a compressive load of 5 MPa at ~400 C up to 9.3x10 25 n/m 2 (E>0.1 MeV.) Following irradiation physical properties were studied to compare the effect of graphite irradiation on microstructure developed under compression and in stress-free condition. Properties included: dimensional change, thermal conductivity, dynamic modulus, and CTE. The effect of stress on open internal porosity was determined through nitrogen adsorption. The IG-110 graphite experienced irradiation-induced creep that is differentiated from irradiation-induced swelling. Irradiation under stress resulted in somewhat greater thermal conductivity and coefficient of thermal expansion. While a significantmore » increase in dynamic modulus occurs, no differentiation between materials irradiated with and without compressive stress was observed. Nitrogen adsorption analysis suggests a difference in pore evolution in the 0.3-40 nm range for graphite irradiated with and without stress, but this evolution is seen to be a small contributor to the overall dimensional change.« less

ICP-MS analysis of fission product diffusion in graphite for High-Temperature Gas-Cooled Reactors

NASA Astrophysics Data System (ADS)

Carter, Lukas M.

Release of radioactive fission products from nuclear fuel during normal reactor operation or in accident scenarios is a fundamental safety concern. Of paramount importance are the understanding and elucidation of mechanisms of chemical interaction, nuclear interaction, and transport phenomena involving fission products. Worldwide efforts to reduce fossil fuel dependence coupled with an increasing overall energy demand have generated renewed enthusiasm toward nuclear power technologies, and as such, these mechanisms continue to be the subjects of vigorous research. High-Temperature Gas-Cooled Reactors (HTGRs or VHTRs) remain one of the most promising candidates for the next generation of nuclear power reactors. An extant knowledge gap specific to HTGR technology derives from an incomplete understanding of fission product transport in major core materials under HTGR operational conditions. Our specific interest in the current work is diffusion in reactor graphite. Development of methods for analysis of diffusion of multiple fission products is key to providing accurate models for fission product release from HTGR core components and the reactor as a whole. In the present work, a specialized diffusion cell has been developed and constructed to facilitate real-time diffusion measurements via ICP-MS. The cell utilizes a helium gas-jet system which transports diffusing fission products to the mass spectrometer using carbon nanoparticles. The setup was designed to replicate conditions present in a functioning HTGR, and can be configured for real-time release or permeation measurements of single or multiple fission products from graphite or other core materials. In the present work, we have analyzed release rates of cesium in graphite grades IG-110, NBG-18, and a commercial grade of graphite, as well as release of iodine in IG-110. Additionally we have investigated infusion of graphite samples with Cs, I, Sr, Ag, and other surrogate fission products for use in release or profile measurements of diffusion coefficients.

Recapturing Graphite-Based Fuel Element Technology for Nuclear Thermal Propulsion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trammell, Michael P; Jolly, Brian C; Miller, James Henry

ORNL is currently recapturing graphite based fuel forms for Nuclear Thermal Propulsion (NTP). This effort involves research and development on materials selection, extrusion, and coating processes to produce fuel elements representative of historical ROVER and NERVA fuel. Initially, lab scale specimens were fabricated using surrogate oxides to develop processing parameters that could be applied to full length NTP fuel elements. Progress toward understanding the effect of these processing parameters on surrogate fuel microstructure is presented.

NUCLEAR REACTOR ELEMENT

DOEpatents

Sanz, M.C.; Scully, C.N.

1961-06-27

The patented fuel element is a hexagonal graphite body having an axial channel therethrough. The graphite is impregnated with uranium which is concentrated near the axial channel. Layers of tantalum nitride and tantalum carbide are disposed on the surface of the body confronting the channel.

Beam impingement angle effects on secondary electron emission characteristics of textured pyrolytic graphite

NASA Technical Reports Server (NTRS)

Curren, A. N.; Jensen, K. A.

1984-01-01

Experimentally determined values of true secondary electron emission and relative values of reflected primary electron yield for untreated and ion-textured pyrolytic graphite over a range of primary electron energy levels and electron beam impingement angles are presented. Information required to develop high efficiency multistage depressed collectors (MDC's) for microwave amplifier traveling-wave tubes for space communication and aircraft applications is provided. To attain the highest possible MDC efficiencies, the electrode surfaces must have low secondary electron emission characteristics. Pyrolytic graphite, a chemically vapor-deposited material, is a particularly promising candidate for this application. The pyrolytic graphite surfaces studied were tested over a range of primary electron beam energies and beam impingement angles from 200 to 2000 eV and direct (0 deg) to near-grazing angles (85 deg), respectively. Surfaces both parallel to and normal to the planes of material deposition were examined. The true secondary electron emission and reflected primary electron yield characteristics of the pyrolytic graphite surfaces are compared to those of sooted control surfaces.

Monovacancy paramagnetism in neutron-irradiated graphite probed by 13C NMR

NASA Astrophysics Data System (ADS)

Zhang, Z. T.; Xu, C.; Dmytriieva, D.; Molatta, S.; Wosnitza, J.; Wang, Y. T.; Helm, M.; Zhou, Shengqiang; Kühne, H.

2017-11-01

We report on the magnetic properties of monovacancy defects in neutron-irradiated graphite, probed by 13C nuclear magnetic resonance spectroscopy. The bulk paramagnetism of the defect moments is revealed by the temperature dependence of the NMR frequency shift and spectral linewidth, both of which follow a Curie behavior, in agreement with measurements of the macroscopic magnetization. Compared to pristine graphite, the fluctuating hyperfine fields generated by the defect moments lead to an enhancement of the 13C nuclear spin-lattice relaxation rate 1/T1 by about two orders of magnitude. With an applied magnetic field of 7.1 T, the temperature dependence of 1/T1 below about 10 K can well be described by a thermally activated form, \

Structure-Property Relationships in Surface-Modified Ceramics. NATO advanced Science Institutes, Series E: Applied Sciences, Volume 170

DTIC Science & Technology

1989-01-01

channelling and scanning electron microscopy (SEM) of highly oriented pyrolytic graphite ( HOPG ), comparative scratch testing results and some ideas on...electrode graphite , HOPG and carbon fibers also show enhanced wear resistance followoing irradiation (6), the extent of which depends upon the initial...literature dealing with damage effects and physical property changes following neutron irradiation of graphite (single and polycrystalline ) in nuclear

Damage, crack growth and fracture characteristics of nuclear grade graphite using the Double Torsion technique

NASA Astrophysics Data System (ADS)

Becker, T. H.; Marrow, T. J.; Tait, R. B.

2011-07-01

The crack initiation and propagation characteristics of two medium grained polygranular graphites, nuclear block graphite (NBG10) and Gilsocarbon (GCMB grade) graphite, have been studied using the Double Torsion (DT) technique. The DT technique allows stable crack propagation and easy crack tip observation of such brittle materials. The linear elastic fracture mechanics (LEFM) methodology of the DT technique was adapted for elastic-plastic fracture mechanics (EPFM) in conjunction with a methodology for directly calculating the J-integral from in-plane displacement fields (JMAN) to account for the non-linearity of graphite deformation. The full field surface displacement measurement techniques of electronic speckle pattern interferometry (ESPI) and digital image correlation (DIC) were used to observe and measure crack initiation and propagation. Significant micro-cracking in the fracture process zone (FPZ) was observed as well as crack bridging in the wake of the crack tip. The R-curve behaviour was measured to determine the critical J-integral for crack propagation in both materials. Micro-cracks tended to nucleate at pores, causing deflection of the crack path. Rising R-curve behaviour was observed, which is attributed to the formation of the FPZ, while crack bridging and distributed micro-cracks are responsible for the increase in fracture resistance. Each contributes around 50% of the irreversible energy dissipation in both graphites.

Multidisciplinary Simulation of Graphite-Composite and Cermet Fuel Elements for NTP Point of Departure Designs

NASA Technical Reports Server (NTRS)

Stewart, Mark E.; Schnitzler, Bruce G.

2015-01-01

This paper compares the expected performance of two Nuclear Thermal Propulsion fuel types. High fidelity, fluid/thermal/structural + neutronic simulations help predict the performance of graphite-composite and cermet fuel types from point of departure engine designs from the Nuclear Thermal Propulsion project. Materials and nuclear reactivity issues are reviewed for each fuel type. Thermal/structural simulations predict thermal stresses in the fuel and thermal expansion mis-match stresses in the coatings. Fluid/thermal/structural/neutronic simulations provide predictions for full fuel elements. Although NTP engines will utilize many existing chemical engine components and technologies, nuclear fuel elements are a less developed engine component and introduce design uncertainty. Consequently, these fuel element simulations provide important insights into NTP engine performance.

Modeling Fission Product Sorption in Graphite Structures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Szlufarska, Izabela; Morgan, Dane; Allen, Todd

2013-04-08

The goal of this project is to determine changes in adsorption and desorption of fission products to/from nuclear-grade graphite in response to a changing chemical environment. First, the project team will employ principle calculations and thermodynamic analysis to predict stability of fission products on graphite in the presence of structural defects commonly observed in very high- temperature reactor (VHTR) graphites. Desorption rates will be determined as a function of partial pressure of oxygen and iodine, relative humidity, and temperature. They will then carry out experimental characterization to determine the statistical distribution of structural features. This structural information will yield distributionsmore » 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 on each type of graphite site. The model will include multiple simultaneous adsorbing species, which will allow for competitive adsorption effects between different fission product species and O and OH (for modeling accident conditions).« less

Beyond the classical kinetic model for chronic graphite oxidation by moisture in high temperature gas-cooled reactors

DOE PAGES

Contescu, Cristian I.; Mee, Robert W.; Lee, Yoonjo; ...

2017-11-03

Four grades of nuclear graphite with various microstructures were subjected to accelerated oxidation tests in helium with traces of moisture and hydrogen in order to evaluate the effects of chronic oxidation on graphite components in high temperature gas cooled reactors. Kinetic analysis showed that the Langmuir-Hinshelwood (LH) model cannot consistently reproduce all results. In particular, at high temperatures and water partial pressures oxidation was always faster than the LH model predicts, with stronger deviations for superfine grain graphite than for medium grain grades. It was also found empirically that the apparent reaction order for water has a sigmoid-type variation withmore » temperature which follows the integral Boltzmann distribution function. This suggests that the apparent activation with temperature of graphite reactive sites that causes deviations from the LH model is rooted in specific structural and electronic properties of surface sites on graphite. A semi-global kinetic model was proposed, whereby the classical LH model was modified with a temperature-dependent reaction order for water. The new Boltzmann-enhanced model (BLH) was shown to consistently predict experimental oxidation rates over large ranges of temperature (800-1100 oC) and partial pressures of water (3-1200 Pa) and hydrogen (0-300 Pa), not only for the four grades of graphite but also for the historic grade H-451. The BLH model offers as more reliable input for modeling the chemical environment effects during the life-time operation of new grades of graphite in advanced nuclear reactors operating at high and very high temperatures.« less

Beyond the classical kinetic model for chronic graphite oxidation by moisture in high temperature gas-cooled reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Contescu, Cristian I.; Mee, Robert W.; Lee, Yoonjo

Four grades of nuclear graphite with various microstructures were subjected to accelerated oxidation tests in helium with traces of moisture and hydrogen in order to evaluate the effects of chronic oxidation on graphite components in high temperature gas cooled reactors. Kinetic analysis showed that the Langmuir-Hinshelwood (LH) model cannot consistently reproduce all results. In particular, at high temperatures and water partial pressures oxidation was always faster than the LH model predicts, with stronger deviations for superfine grain graphite than for medium grain grades. It was also found empirically that the apparent reaction order for water has a sigmoid-type variation withmore » temperature which follows the integral Boltzmann distribution function. This suggests that the apparent activation with temperature of graphite reactive sites that causes deviations from the LH model is rooted in specific structural and electronic properties of surface sites on graphite. A semi-global kinetic model was proposed, whereby the classical LH model was modified with a temperature-dependent reaction order for water. The new Boltzmann-enhanced model (BLH) was shown to consistently predict experimental oxidation rates over large ranges of temperature (800-1100 oC) and partial pressures of water (3-1200 Pa) and hydrogen (0-300 Pa), not only for the four grades of graphite but also for the historic grade H-451. The BLH model offers as more reliable input for modeling the chemical environment effects during the life-time operation of new grades of graphite in advanced nuclear reactors operating at high and very high temperatures.« less

A phase-field approach to model multi-axial and microstructure dependent fracture in nuclear grade graphite

DOE PAGES

Chakraborty, Pritam; Sabharwall, Piyush; Carroll, Mark C.

2016-04-07

The fracture behavior of nuclear grade graphites is strongly influenced by underlying microstructural features such as the character of filler particles, and the distribution of pores and voids. These microstructural features influence the crack nucleation and propagation behavior, resulting in quasi-brittle fracture with a tortuous crack path and significant scatter in measured bulk strength. This paper uses a phase-field method to model the microstructural and multi-axial fracture in H-451, a historic variant of nuclear graphite that provides the basis for an idealized study on a legacy grade. The representative volume elements are constructed from randomly located pores with random sizemore » obtained from experimentally determined log-normal distribution. The representative volume elements are then subjected to simulated multi-axial loading, and a reasonable agreement of the resulting fracture stress with experiments is obtained. Finally, quasi-brittle stress-strain evolution with a tortuous crack path is also observed from the simulations and is consistent with experimental results.« less

REACTOR HAVING NaK-UO$sub 2$ SLURRY HELICALLY POSITIONED IN A GRAPHITE MODERATOR

DOEpatents

Rodin, M.B.; Carter, J.C.

1962-05-15

A reactor utilizing 20% enriched uranium consists of a central graphite island in cylindrical form, with a spiral coil of tubing fitting against the central island. An external graphite moderator is placed around the central island and coil. A slurry of uranium dioxide dispersed in alkali metal passes through the coil to transfer heat externally to the reactor. There are also conventional controls for regulating the nuclear reaction. (AEC)

Preliminary Results of IS Plasma Focus as a Breeder of Short-Lived Radioisotopes 12C(d,n)13N

NASA Astrophysics Data System (ADS)

Sadat Kiai, S. M.; Elahi, M.; Adlparvar, S.; Shahhoseini, E.; Sheibani, S.; Ranjber akivaj, H.; Alhooie, S.; Safarien, A.; Farhangi, S.; Aghaei, N.; Amini, S.; Khalaj, M. M.; Zirak, A. R.; Dabirzadeh, A. A.; Soleimani, J.; Torkzadeh, F.; Mousazadeh, M. M.; Moradi, K.; Abdollahzadeh, M.; Talaei, A.; Zaeem, A. A.; Moslehi, A.; Kashani, A.; Babazadeh, A. R.; Bagiyan, F.; Ardestani, M.; Roozbahani, A.; Pourbeigi, H.; Tajik Ahmadi, H.; Ahmadifaghih, M. A.; Mahlooji, M. S.; Mortazavi, B. N.; Zahedi, F.

2011-04-01

Modified IS (Iranian Sun) plasma focus (10 kJ,15 kV, 94 μF, 0.1 Hz) has been used to produce the short-lived radioisotope 13N (half-life of 9.97 min) through 12C(d,n)13N nuclear reaction. The filling gas was 1.5-3 torr of hydrogen (60%) deuterium (40%) mixture. The target was solid nuclear grade graphite with 5 mm thick, 9 cm width and 13 in length. The activations of the exogenous target on average of 20 shots (only one-third acceptable) through 10-13 kV produced the 511 keV gamma rays. Another peak found at the 570 keV gamma of which both was measured by a NaI portable gamma spectrometer calibrated by a 137Cs 0.25 μCi sealed reference source with its single line at 661.65 keV and 22Na 0.1 μCi at 511 keV. To measure the gamma rays, the graphite target converts to three different phases; solid graphite, powder graphite, and powder graphite in water solution. The later phase approximately has a doubled activity with respect to the solid graphite target up to 0.5 μCi of 511 keV and 1.1 μCi of 570 keV gamma lines were produced. This increment in activity was perhaps due to structural transformation of graphite powder to nano-particles characteristic in liquid water.

Large-Scale Weibull Analysis of H-451 Nuclear- Grade Graphite Specimen Rupture Data

NASA Technical Reports Server (NTRS)

Nemeth, Noel N.; Walker, Andrew; Baker, Eric H.; Murthy, Pappu L.; Bratton, Robert L.

2012-01-01

A Weibull analysis was performed of the strength distribution and size effects for 2000 specimens of H-451 nuclear-grade graphite. The data, generated elsewhere, measured the tensile and four-point-flexure room-temperature rupture strength of specimens excised from a single extruded graphite log. Strength variation was compared with specimen location, size, and orientation relative to the parent body. In our study, data were progressively and extensively pooled into larger data sets to discriminate overall trends from local variations and to investigate the strength distribution. The CARES/Life and WeibPar codes were used to investigate issues regarding the size effect, Weibull parameter consistency, and nonlinear stress-strain response. Overall, the Weibull distribution described the behavior of the pooled data very well. However, the issue regarding the smaller-than-expected size effect remained. This exercise illustrated that a conservative approach using a two-parameter Weibull distribution is best for designing graphite components with low probability of failure for the in-core structures in the proposed Generation IV (Gen IV) high-temperature gas-cooled nuclear reactors. This exercise also demonstrated the continuing need to better understand the mechanisms driving stochastic strength response. Extensive appendixes are provided with this report to show all aspects of the rupture data and analytical results.

New insights into canted spiro carbon interstitial in graphite

NASA Astrophysics Data System (ADS)

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.

A (13)C NMR analysis of the effects of electron radiation on graphite/polyetherimide composites

NASA Technical Reports Server (NTRS)

Ferguson, Milton W.

1989-01-01

Initial investigations have been made into the use of high resolution nuclear magnetic resonance (NMR) for the characterization of radiation effects in graphite and Kevlar fibers, polymers, and the fiber/matrix interface in graphite/polyetherimide composites. Sample preparation techniques were refined. Essential equipment has been procured. A new NMR probe was constructed to increase the proton signal-to-noise ratio. Problem areas have been identified and plans developed to resolve them.

Initial Assessment of X-Ray Computer Tomography image analysis for material defect microstructure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kane, Joshua James; Windes, William Enoch

2016-06-01

The original development work leading to this report was focused on the non destructive three-dimensional (3-D) characterization of nuclear graphite as a means to better understand the nature of the inherent pore structure. The pore structure of graphite and its evolution under various environmental factors such as irradiation, mechanical stress, and oxidation plays an important role in their observed properties and characteristics. If we are to transition from an empirical understanding of graphite behavior to a truly predictive mechanistic understanding the pore structure must be well characterized and understood. As the pore structure within nuclear graphite is highly interconnected andmore » truly 3-D in nature, 3-D characterization techniques are critical. While 3-D characterization has been an excellent tool for graphite pore characterization, it is applicable to a broad number of materials systems over many length scales. Given the wide range of applications and the highly quantitative nature of the tool, it is quite surprising to discover how few materials researchers understand and how valuable of a tool 3-D image processing and analysis can be. Ultimately, this report is intended to encourage broader use of 3 D image processing and analysis in materials science and engineering applications, more specifically nuclear-related materials applications, by providing interested readers with enough familiarity to explore its vast potential in identifying microstructure changes. To encourage this broader use, the report is divided into two main sections. Section 2 provides an overview of some of the key principals and concepts needed to extract a wide variety of quantitative metrics from a 3-D representation of a material microstructure. The discussion includes a brief overview of segmentation methods, connective components, morphological operations, distance transforms, and skeletonization. Section 3 focuses on the application of concepts from Section 2 to relevant materials at Idaho National Laboratory. In this section, image analysis examples featuring nuclear graphite will be discussed in detail. Additionally, example analyses from Transient Reactor Test Facility low-enriched uranium conversion, Advanced Gas Reactor like compacts, and tristructural isotopic particles are shown to give a broader perspective of the applicability to relevant materials of interest.« less

A probabilisitic based failure model for components fabricated from anisotropic graphite

NASA Astrophysics Data System (ADS)

Xiao, Chengfeng

The nuclear moderator for high temperature nuclear reactors are fabricated from graphite. During reactor operations graphite components are subjected to complex stress states arising from structural loads, thermal gradients, neutron irradiation damage, and seismic events. Graphite is a quasi-brittle material. Two aspects of nuclear grade graphite, i.e., material anisotropy and different behavior in tension and compression, are explicitly accounted for in this effort. Fracture mechanic methods are useful for metal alloys, but they are problematic for anisotropic materials with a microstructure that makes it difficult to identify a "critical" flaw. In fact cracking in a graphite core component does not necessarily result in the loss of integrity of a nuclear graphite core assembly. A phenomenological failure criterion that does not rely on flaw detection has been derived that accounts for the material behaviors mentioned. The probability of failure of components fabricated from graphite is governed by the scatter in strength. The design protocols being proposed by international code agencies recognize that design and analysis of reactor core components must be based upon probabilistic principles. The reliability models proposed herein for isotropic graphite and graphite that can be characterized as being transversely isotropic are another set of design tools for the next generation very high temperature reactors (VHTR) as well as molten salt reactors. The work begins with a review of phenomenologically based deterministic failure criteria. A number of this genre of failure models are compared with recent multiaxial nuclear grade failure data. Aspects in each are shown to be lacking. The basic behavior of different failure strengths in tension and compression is exhibited by failure models derived for concrete, but attempts to extend these concrete models to anisotropy were unsuccessful. The phenomenological models are directly dependent on stress invariants. A set of invariants, known as an integrity basis, was developed for a non-linear elastic constitutive model. This integrity basis allowed the non-linear constitutive model to exhibit different behavior in tension and compression and moreover, the integrity basis was amenable to being augmented and extended to anisotropic behavior. This integrity basis served as the starting point in developing both an isotropic reliability model and a reliability model for transversely isotropic materials. At the heart of the reliability models is a failure function very similar in nature to the yield functions found in classic plasticity theory. The failure function is derived and presented in the context of a multiaxial stress space. States of stress inside the failure envelope denote safe operating states. States of stress on or outside the failure envelope denote failure. The phenomenological strength parameters associated with the failure function are treated as random variables. There is a wealth of failure data in the literature that supports this notion. The mathematical integration of a joint probability density function that is dependent on the random strength variables over the safe operating domain defined by the failure function provides a way to compute the reliability of a state of stress in a graphite core component fabricated from graphite. The evaluation of the integral providing the reliability associated with an operational stress state can only be carried out using a numerical method. Monte Carlo simulation with importance sampling was selected to make these calculations. The derivation of the isotropic reliability model and the extension of the reliability model to anisotropy are provided in full detail. Model parameters are cast in terms of strength parameters that can (and have been) characterized by multiaxial failure tests. Comparisons of model predictions with failure data is made and a brief comparison is made to reliability predictions called for in the ASME Boiler and Pressure Vessel Code. Future work is identified that would provide further verification and augmentation of the numerical methods used to evaluate model predictions.

Study of evaporating the irradiated graphite in equilibrium low-temperature plasma

NASA Astrophysics Data System (ADS)

Bespala, E. V.; Novoselov, I. Yu.; Pavlyuk, A. O.; Kotlyarevskiy, S. G.

2018-01-01

The paper describes a problem of accumulation of irradiated graphite due to operation of uranium-graphite nuclear reactors. The main noncarbon contaminants that contribute to the overall activity of graphite elements are iso-topes 137Cs, 60Co, 90Sr, 36Cl, and 3H. A method was developed for processing of irradiated graphite ensuring the volu-metric decontamination of samples. The calculation results are presented for equilibrium composition of plasma-chemical reactions in systems "irradiated graphite-argon" and "irradiated graphite-helium" for a wide range of tem-peratures. The paper describes a developed mathematical model for the process of purification of a porous graphite surface treated by equilibrium low-temperature plasma. The simulation results are presented for the rate of sublimation of radioactive contaminants as a function of plasma temperature and plasma flow velocity when different plasma-forming gases are used. The extraction coefficient for the contaminant 137Cs from the outer side of graphite pores was calculated. The calculations demonstrated the advantages of using a lighter plasma forming gas, i.e., helium.

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

PubMed

Theodosiou, Alex; 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.

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

PubMed Central

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

Electron reflection and secondary emission characteristics of sputter-textured pyrolytic graphite surfaces

NASA Technical Reports Server (NTRS)

Wintucky, E. G.; Curren, A. N.; Sovey, J. S.

1981-01-01

Measurements are presented of secondary electron emission and reflected primary electron characteristics of sputter-textured pyrolitic graphite surfaces with microstructures of various sizes and densities, made with an Auger cylindrical mirror analyzer in a high-vacuum chamber at pressures below 1.33 x 10 to the -7th N/sq m (10 to the -9th torr). A dense, tall, thin, spire-like microstructure, obtained at ion energies of 1000 eV and ion current densities of 5 mA/sq cm, is the most effective. The secondary electron emission from such a surface is lower than that of soot, whose secondary emission is among the lowest of any material. At a primary electron energy of 1000 eV, the secondary electron emission yield of smooth CU is about 350% greater than the lowest value obtained for sputter-textured pyrolitic graphite. The reflected primary electron index of smooth Cu is a factor of 80 greater. If the secondary electron emission yield is reduced to 0.3, which is possible with sputter-textured pyrolitic graphite, the traveling wave tube collector efficiency could be improved by as much as 4% over that for smooth copper.

Temperature Dependence of Power Reflectivity of the First-Wall Materials in the Synchrotron Radiation Range

NASA Astrophysics Data System (ADS)

Takada, Noriharu; Nagatsu, Masaaki; Shimada, Michiya

1995-07-01

The temperature dependence of power reflectivity in the synchrotron radiation range was measured for candidate first-wall materials of the fusion reactor, such as B4C-coated isotropic graphite, C/C composite material, silicon carbide (SiC), tungsten (W), molybdenum (Mo) and SUS-316. The measurements were carried out using a vacuum vessel with a pressure of about 3 mTorr to avoid oxidation. Distinct temperature dependence of reflectivity was observed only for B4C-coated isotropic graphite. For the other materials, power reflectivities were insensitive to temperature in the range from 300 K to ˜900 K. Theoretical analysis of the results is also presented.

Measurement of partonic nuclear effects in deep-inelastic neutrino scattering using MINERvA

NASA Astrophysics Data System (ADS)

Mousseau, J.; Wospakrik, M.; Aliaga, L.; Altinok, O.; Bellantoni, L.; Bercellie, A.; Betancourt, M.; Bodek, A.; Bravar, A.; Budd, H.; Cai, T.; Carneiro, M. F.; Christy, M. E.; Chvojka, J.; da Motta, H.; Devan, J.; Dytman, S. A.; Díaz, G. A.; Eberly, B.; Felix, J.; Fields, L.; Fine, R.; Gago, A. M.; Galindo, R.; Gallagher, H.; Ghosh, A.; Golan, T.; Gran, R.; Harris, D. A.; Higuera, A.; Hurtado, K.; Kiveni, M.; Kleykamp, J.; Kordosky, M.; Le, T.; Maher, E.; Manly, S.; Mann, W. A.; Marshall, C. M.; Martinez Caicedo, D. A.; McFarland, K. S.; McGivern, C. L.; McGowan, A. M.; Messerly, B.; Miller, J.; Mislivec, A.; Morfín, J. G.; Naples, D.; Nelson, J. K.; Norrick, A.; Nuruzzaman; Osta, J.; Paolone, V.; Park, J.; Patrick, C. E.; Perdue, G. N.; Rakotondravohitra, L.; Ramirez, M. A.; Ransome, R. D.; Ray, H.; Ren, L.; Rimal, D.; Rodrigues, P. A.; Ruterbories, D.; Schellman, H.; Schmitz, D. W.; Solano Salinas, C. J.; Tagg, N.; Tice, B. G.; Valencia, E.; Walton, T.; Wolcott, J.; Zavala, G.; Zhang, D.; Minerν A Collaboration

2016-04-01

The MINERvA Collaboration reports a novel study of neutrino-nucleus charged-current deep inelastic scattering (DIS) using the same neutrino beam incident on targets of polystyrene, graphite, iron, and lead. Results are presented as ratios of C, Fe, and Pb to CH. The ratios of total DIS cross sections as a function of neutrino energy and flux-integrated differential cross sections as a function of the Bjorken scaling variable x are presented in the neutrino-energy range of 5-50 GeV. Based on the predictions of charged-lepton scattering ratios, good agreement is found between the data and prediction at medium x and low neutrino energy. However, the ratios appear to be below predictions in the vicinity of the nuclear shadowing region, x <0.1 . This apparent deficit, reflected in the DIS cross-section ratio at high Eν, is consistent with previous MINERvA observations [B. Tice et al. (MINERvA Collaboration), Phys. Rev. Lett. 112, 231801 (2014).] and with the predicted onset of nuclear shadowing with the axial-vector current in neutrino scattering.

Measurement of partonic nuclear effects in deep-inelastic neutrino scattering using MINERvA

DOE PAGES

Mousseau, J.

2016-04-19

Here, the MINERvA Collaboration reports a novel study of neutrino-nucleus charged-current deep inelastic scattering (DIS) using the same neutrino beam incident on targets of polystyrene, graphite, iron, and lead. Results are presented as ratios of C, Fe, and Pb to CH. The ratios of total DIS cross sections as a function of neutrino energy and flux-integrated differential cross sections as a function of the Bjorken scaling variable x are presented in the neutrino-energy range of 5–50 GeV. Based on the predictions of charged-lepton scattering ratios, good agreement is found between the data and prediction at medium x and low neutrino energy.more » However, the ratios appear to be below predictions in the vicinity of the nuclear shadowing region, x < 0.1. This apparent deficit, reflected in the DIS cross-section ratio at high Eν, is consistent with previous MINERvA observations [B. Tice (MINERvA Collaboration), Phys. Rev. Lett. 112, 231801 (2014).] and with the predicted onset of nuclear shadowing with the axial-vector current in neutrino scattering.« less

Advanced Ceramics for Use as Fuel Element Materials in Nuclear Thermal Propulsion Systems

NASA Technical Reports Server (NTRS)

Valentine, Peter G.; Allen, Lee R.; Shapiro, Alan P.

2012-01-01

With the recent start (October 2011) of the joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) Advanced Exploration Systems (AES) Nuclear Cryogenic Propulsion Stage (NCPS) Program, there is renewed interest in developing advanced ceramics for use as fuel element materials in nuclear thermal propulsion (NTP) systems. Three classes of fuel element materials are being considered under the NCPS Program: (a) graphite composites - consisting of coated graphite elements containing uranium carbide (or mixed carbide), (b) cermets (ceramic/metallic composites) - consisting of refractory metal elements containing uranium oxide, and (c) advanced carbides consisting of ceramic elements fabricated from uranium carbide and one or more refractory metal carbides [1]. The current development effort aims to advance the technology originally developed and demonstrated under Project Rover (1955-1973) for the NERVA (Nuclear Engine for Rocket Vehicle Application) [2].

Nuclear fuel elements and method of making same

DOEpatents

Schweitzer, Donald G.

1992-01-01

A nuclear fuel element for a high temperature gas nuclear reactor that has an average operating temperature in excess of 2000.degree. C., and a method of making such a fuel element. The fuel element is characterized by having fissionable fuel material localized and stabilized within pores of a carbon or graphite member by melting the fissionable material to cause it to chemically react with the carbon walls of the pores. The fissionable fuel material is further stabilized and localized within the pores of the graphite member by providing one or more coatings of pyrolytic carbon or diamond surrounding the porous graphite member so that each layer defines a successive barrier against migration of the fissionable fuel from the pores, and so that the outermost layer of pyrolytic carbon or diamond forms a barrier between the fissionable material and the moderating gases used in an associated high temperature gas reactor. The method of the invention provides for making such new elements either as generally spherically elements, or as flexible filaments, or as other relatively small-sized fuel elements that are particularly suited for use in high temperature gas reactors.

Compatibility of the Radio Frequency Mass Gauge with Graphite-Epoxy Composite Tanks

NASA Technical Reports Server (NTRS)

Zimmerli, G. A.; Mueller, C. H.

2015-01-01

The radio frequency mass gauge (RFMG) is a low-gravity propellant quantity gauge being developed at NASA for possible use in long-duration space missions utilizing cryogenic propellants. As part of the RFMG technology development process, we evaluated the compatibility of the RFMG with a graphite-epoxy composite material used to construct propellant tanks. The key material property that can affect compatibility with the RFMG is the electrical conductivity. Using samples of 8552/IM7 graphite-epoxy composite, we characterized the resistivity and reflectivity over a range of frequencies. An RF impedance analyzer was used to characterize the out-of-plane electrical properties (along the sample thickness) in the frequency range 10 to 1800 MHZ. The resistivity value at 500 MHz was 4.8 ohm-cm. Microwave waveguide measurements of samples in the range 1.7 - 2.6 GHz, performed by inserting the samples into a WR-430 waveguide, showed reflectivity values above 98%. Together, these results suggested that a tank constructed from graphite/epoxy composite would produce good quality electromagnetic tank modes, which is needed for the RFMG. This was verified by room-temperature measurements of the electromagnetic modes of a 2.4 m diameter tank constructed by Boeing from similar graphite-epoxy composite material. The quality factor Q of the tank electromagnetic modes, measured via RF reflection measurements from an antenna mounted in the tank, was typically in the range 400 less than Q less than 3000. The good quality modes observed in the tank indicate that the RFMG is compatible with graphite-epoxy tanks, and thus the RFMG could be used as a low-gravity propellant quantity gauge in such tanks filled with cryogenic propellants.

High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mcwilliams, A. J.

2015-09-08

This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniquesmore » through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.« less

Design of Modern Reactors for Synthesis of Thermally Expanded Graphite.

PubMed

Strativnov, Eugene V

2015-12-01

One of the most progressive trends in the development of modern science and technology is the creation of energy-efficient technologies for the synthesis of nanomaterials. Nanolayered graphite (thermally exfoliated graphite) is one of the key important nanomaterials of carbon origin. Due to its unique properties (chemical and thermal stability, ability to form without a binder, elasticity, etc.), it can be used as an effective absorber of organic substances and a material for seal manufacturing for such important industries as gas transportation and automobile. Thermally expanded graphite is a promising material for the hydrogen and nuclear energy industries. The development of thermally expanded graphite production is resisted by high specific energy consumption during its manufacturing and by some technological difficulties. Therefore, the creation of energy-efficient technology for its production is very promising.

Measurement of leakage neutron spectra from graphite cylinders irradiated with D-T neutrons for validation of evaluated nuclear data.

PubMed

Luo, F; Han, R; Chen, Z; Nie, Y; Shi, F; Zhang, S; Lin, W; Ren, P; Tian, G; Sun, Q; Gou, B; Ruan, X; Ren, J; Ye, M

2016-10-01

A benchmark experiment for validation of graphite data evaluated from nuclear data libraries was conducted for 14MeV neutrons irradiated on graphite cylinder samples. The experiments were performed using the benchmark experimental facility at the China Institute of Atomic Energy (CIAE). The leakage neutron spectra from the surface of graphite (Φ13cm×20cm) at 60° and 120° and graphite (Φ13cm×2cm) at 60° were measured by the time-of-flight (TOF) method. The obtained results were compared with the measurements made by the Monte Carlo neutron transport code MCNP-4C with the ENDF/B-VII.1, CENDL-3.1 and JENDL-4.0 libraries. The results obtained from a 20cm-thick sample revealed that the calculation results with CENDL-3.1 and JENDL-4.0 libraries showed good agreements with the experiments conducted in the whole energy region. However, a large discrepancy of approximately 40% was observed below the 3MeV energy region with the ENDF/B-VII.1 library. For the 2cm-thick sample, the calculated results obtained from the abovementioned three libraries could not reproduce the experimental data in the energy range of 5-7MeV. The graphite data in CENDL-3.1 were verified for the first time and were proved to be reliable. Copyright © 2016 Elsevier Ltd. All rights reserved.

Approaches to Deal with Irradiated Graphite in Russia - Proposal for New IAEA CRP on Graphite Waste Management - 12364

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kascheev, Vladimir; Poluektov, Pavel; Ustinov, Oleg

The problems of spent reactor graphite are being shown, the options of its disposal is considered. Burning method is selected as the most efficient and waste-free. It is made a comparison of amounts of {sup 14}C that entering the environment in a natural way during the operation of nuclear power plants (NPPs) and as a result of the proposed burning of spent reactor graphite. It is shown the possibility of burning graphite with the arrival of {sup 14}C into the atmosphere within the maximum allowable emissions. This paper analyzes the different ways of spent reactor graphite treatment. It is shownmore » the possibility of its reprocessing by burning method in the air flow. It is estimated the effect of this technology to the overall radiation environment and compared its contribution to the general background radiation due to cosmic radiation and NPPs emission. It is estimated the maximum permissible speeds of burning reactor graphite (for example, RBMK graphite) for areas with different conditions of agricultural activities. (authors)« less

Chemical Characterization and Removal of Carbon-14 from Irradiated Graphite II - 13023

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dunzik-Gougar, Mary Lou; Cleaver, James; LaBrier, Daniel

2013-07-01

Approximately 250,000 tonnes of irradiated graphite waste exists worldwide and that quantity is expected to increase with decommissioning of Generation II reactors and deployment of Generation IV gas-cooled, graphite moderated reactors. This situation indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 (C-14), with a half-life of 5730 years. Study of irradiated graphite from some nuclear reactors indicates C-14 is concentrated on the outer 5 mm of the graphite structure. The aim of the research presented last year and updated here is to identify the chemical form of C-14more » in irradiated graphite and develop a practical method by which C-14 can be removed. A nuclear-grade graphite, NBG-18, and a high-surface-area graphite foam, POCOFoam{sup R}, were exposed to liquid nitrogen (to increase the quantity of C-14 precursor) and neutron-irradiated (10{sup 13} neutrons/cm{sup 2}/s). Finer grained NBG-25 was not exposed to liquid nitrogen prior to irradiation at a neutron flux on the order of 10{sup 14} /cm{sup 2}/s. Characterization of pre- and post-irradiation graphite was conducted to determine the chemical environment and quantity of C-14 and its precursors via the use of surface sensitive characterization techniques. Scanning Electron Microscopy (SEM) was used to evaluate the morphological features of graphite samples. The concentration, chemical composition, and bonding characteristics of C-14 and its precursors were determined through X-ray Photoelectron Spectroscopy (XPS), Time-of-Flight Secondary Ion Mass Spectrometry (SIMS), and Energy Dispersive X-ray Analysis Spectroscopy (EDX). Results of post-irradiation characterization of these materials indicate a variety of surface functional groups containing carbon, oxygen, nitrogen and hydrogen. During thermal treatment, irradiated graphite samples are heated in the presence of an inert carrier gas (with or without oxidant gas), which carries off gaseous products released during treatment. Graphite gasification occurs via interaction with adsorbed oxygen complexes. Experiments in argon were performed at 900 deg. C and 1400 deg. C to evaluate the selective removal of C-14. Thermal treatment also was performed with the addition of 3 and 5 volume % oxygen at temperatures 700 deg. C and 1400 deg. C. Thermal treatment experiments were evaluated for the effective selective removal of C-14. Lower temperatures and oxygen levels correlated to more efficient C-14 removal. (authors)« less

New insight into UO 2F 2 particulate structure by micro-Raman spectroscopy

DOE PAGES

Stefaniak, Elzbieta A.; Darchuk, Larysa; Sapundjiev, Danislav; ...

2013-02-19

Uranyl fluoride particles produced via hydrolysis of uranium hexafluoride have been deposited on different substrates: polished graphite disks, silver foil, stainless steel and gold-coated silicon wafer, and measured with micro-Raman spectroscopy (MRS). All three metallic substrates enhanced the Raman signal delivered by UO 2F 2 in comparison to graphite. The fundamental stretching of the U–O band appeared at 867 cm –1 in case of the graphite substrate, while in case of the others it was shifted to lower frequencies (down to 839 cm –1). All applied metallic substrates showed the expected effect of Raman signal enhancement; however the gold layermore » appeared to be most effective. Lastly, application of new substrates provides more information on the molecular structure of uranyl fluoride precipitation, which is interesting for nuclear safeguards and nuclear environmental analysis.« less

Monovacancy paramagnetism in neutron-irradiated graphite probed by 13C NMR.

PubMed

Zhang, Z T; Xu, C; Dmytriieva, D; Molatta, S; Wosnitza, J; Wang, Y T; Helm, M; Zhou, Shengqiang; Kühne, H

2017-10-20

We report on the magnetic properties of monovacancy defects in neutron-irradiated graphite, probed by 13 C nuclear magnetic resonance spectroscopy. The bulk paramagnetism of the defect moments is revealed by the temperature dependence of the NMR frequency shift and spectral linewidth, both of which follow a Curie behavior, in agreement with measurements of the macroscopic magnetization. Compared to pristine graphite, the fluctuating hyperfine fields generated by the defect moments lead to an enhancement of the 13 C nuclear spin-lattice relaxation rate [Formula: see text] by about two orders of magnitude. With an applied magnetic field of 7.1 T, the temperature dependence of [Formula: see text] below about 10 K can well be described by a thermally activated form, [Formula: see text], yielding a singular Zeeman energy of ([Formula: see text]) meV, in excellent agreement with the sole presence of polarized, non-interacting defect moments.

Property changes of G347A graphite due to neutron irradiation

DOE PAGES

Campbell, Anne A.; Katoh, Yutai; Snead, Mary A.; ...

2016-08-18

A new, fine-grain nuclear graphite, grade G347A from Tokai Carbon Co., Ltd., has been irradiated in the High Flux Isotope Reactor at Oak Ridge National Laboratory to study the materials property changes that occur when exposed to neutron irradiation at temperatures of interest for Generation-IV nuclear reactor applications. Specimen temperatures ranged from 290°C to 800 °C with a maximum neutron fluence of 40 × 10 25 n/m 2 [E > 0.1 MeV] (~30dpa). Lastly, observed behaviors include: anisotropic behavior of dimensional change in an isotropic graphite, Young's modulus showing parabolic fluence dependence, electrical resistivity increasing at low fluence and additionalmore » increase at high fluence, thermal conductivity rapidly decreasing at low fluence followed by continued degradation, and a similar plateau value of the mean coefficient of thermal expansion for all irradiation temperatures.« less

REFLECTOR FOR NEUTRONIC REACTORS

DOEpatents

Fraas, A.P.

1963-08-01

A reflector for nuclear reactors that comprises an assembly of closely packed graphite rods disposed with their major axes substantially perpendicular to the interface between the reactor core and the reflector is described. Each graphite rod is round in transverse cross section at (at least) its interface end and is provided, at that end, with a coaxial, inwardly tapering hole. (AEC)

Lighting Studies for Fuelling Machine Deployed Visual Inspection Tool

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stoots, Carl; Griffith, George

2015-04-01

Under subcontract to James Fisher Nuclear, Ltd., INL has been reviewing advanced vision systems for inspection of graphite in high radiation, high temperature, and high pressure environments. INL has performed calculations and proof-of-principle measurements of optics and lighting techniques to be considered for visual inspection of graphite fuel channels in AGR reactors in UK.

3. VIEW OF ADDITION TO BUILDING 444. IN THE MID ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

3. VIEW OF ADDITION TO BUILDING 444. IN THE MID 1950s, RADIOGRAPHY VAULTS, A GRAPHITE STORAGE AND CUTTING AREA, AND A GRAPHITE PRODUCTION PROCESSING AREA WERE ADDED TO BUILDING 444. (1956) - Rocky Flats Plant, Non-Nuclear Production Facility, South of Cottonwood Avenue, west of Seventh Avenue & east of Building 460, Golden, Jefferson County, CO

Thermal Properties of G-348 Graphite

DOE Office of Scientific and Technical Information (OSTI.GOV)

McEligot, Donald; Swank, W. David; Cottle, David L.

2016-05-01

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

Thermal Properties of G-348 Graphite

DOE Office of Scientific and Technical Information (OSTI.GOV)

McEligot, Donald M.; Swank, W. David; Cottle, David L.

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

Alloying of steel and graphite by hydrogen in nuclear reactor

NASA Astrophysics Data System (ADS)

Krasikov, E.

2017-02-01

In traditional power engineering hydrogen may be one of the first primary source of equipment damage. This problem has high actuality for both nuclear and thermonuclear power engineering. Study of radiation-hydrogen embrittlement of the steel raises the question concerning the unknown source of hydrogen in reactors. Later unexpectedly high hydrogen concentrations were detected in irradiated graphite. It is necessary to look for this source of hydrogen especially because hydrogen flakes were detected in reactor vessels of Belgian NPPs. As a possible initial hypothesis about the enigmatical source of hydrogen one can propose protons generation during beta-decay of free neutrons поскольку inasmuch as protons detected by researches at nuclear reactors as witness of beta-decay of free neutrons.

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

PubMed

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

2018-03-13

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

Nuclear Rocket Technology Conference

NASA Technical Reports Server (NTRS)

1966-01-01

The Lewis Research Center has a strong interest in nuclear rocket propulsion and provides active support of the graphite reactor program in such nonnuclear areas as cryogenics, two-phase flow, propellant heating, fluid systems, heat transfer, nozzle cooling, nozzle design, pumps, turbines, and startup and control problems. A parallel effort has also been expended to evaluate the engineering feasibility of a nuclear rocket reactor using tungsten-matrix fuel elements and water as the moderator. Both of these efforts have resulted in significant contributions to nuclear rocket technology. Many successful static firings of nuclear rockets have been made with graphite-core reactors. Sufficient information has also been accumulated to permit a reasonable Judgment as to the feasibility of the tungsten water-moderated reactor concept. We therefore consider that this technoIogy conference on the nuclear rocket work that has been sponsored by the Lewis Research Center is timely. The conference has been prepared by NASA personnel, but the information presented includes substantial contributions from both NASA and AEC contractors. The conference excludes from consideration the many possible mission requirements for nuclear rockets. Also excluded is the direct comparison of nuclear rocket types with each other or with other modes of propulsion. The graphite reactor support work presented on the first day of the conference was partly inspired through a close cooperative effort between the Cleveland extension of the Space Nuclear Propulsion Office (headed by Robert W. Schroeder) and the Lewis Research Center. Much of this effort was supervised by Mr. John C. Sanders, chairman for the first day of the conference, and by Mr. Hugh M. Henneberry. The tungsten water-moderated reactor concept was initiated at Lewis by Mr. Frank E. Rom and his coworkers. The supervision of the recent engineering studies has been shared by Mr. Samuel J. Kaufman, chairman for the second day of the conference, and Mr. Roy V. Humble. Dr. John C. Eward served as general chairman for the conference.

Effects of carbon/graphite fiber contamination on high voltage electrical insulation

NASA Technical Reports Server (NTRS)

Garrity, T.; Eichler, C.

1980-01-01

The contamination mechanics and resulting failure modes of high voltage electrical insulation due to carbon/graphite fibers were examined. The high voltage insulation vulnerability to carbon/graphite fiber induced failure was evaluated using a contamination system which consisted of a fiber chopper, dispersal chamber, a contamination chamber, and air ducts and suction blower. Tests were conducted to evaluate the effects of fiber length, weathering, and wetness on the insulator's resistance to carbon/graphite fibers. The ability of nuclear, fossil, and hydro power generating stations to maintain normal power generation when the surrounding environment is contaminated by an accidental carbon fiber release was investigated. The vulnerability assessment included only the power plant generating equipment and its associated controls, instrumentation, and auxiliary and support systems.

Microwave limb sounder, graphite epoxy support structure

NASA Technical Reports Server (NTRS)

Pynchon, G.

1980-01-01

The manufacturing and processing procedures which were used to fabricate a precision graphite/epoxy support structure for a spherical microwave reflecting surface are described. The structure was made fromm GY-70/930 ultra high modulus graphite prepreg, laminated to achieve an isotropic in plane thermal expansion of less than + or - 0.1 PPM/F. The structure was hand assembled to match the interface of the reflective surface, which was an array of 18 flexure supported, aluminum, spherically contoured tiles. Structural adhesives were used in the final assembly to bond the elements into their final configuration. A eutectic metal coating was applied to the composite surface to reduce dimensional instabilities arising from changes in the composite epoxy moisture content due to environmental effects. Basic materials properties data are reported and the results of a finite element structural analysis are referenced.

Analyzing the impact of reactive transport on the repository performance of TRISO fuel

NASA Astrophysics Data System (ADS)

Schmidt, Gregory

One of the largest determiners of the amount of electricity generated by current nuclear reactors is the efficiency of the thermodynamic cycle used for power generation. Current light water reactors (LWR) have an efficiency of 35% or less for the conversion of heat energy generated by the reactor to electrical energy. If this efficiency could be improved, more power could be generated from equivalent volumes of nuclear fuel. One method of improving this efficiency is to use a coolant flow that operates at a much higher temperature for electricity production. A reactor design that is currently proposed to take advantage of this efficiency is a graphite-moderated, helium-cooled reactor known as a High Temperature Gas Reactor (HTGR). There are significant differences between current LWR's and the proposed HTGR's but most especially in the composition of the nuclear fuel. For LWR's, the fuel elements consist of pellets of uranium dioxide or plutonium dioxide that are placed in long tubes made of zirconium metal alloys. For HTGR's, the fuel, known as TRISO (TRIstructural-ISOtropic) fuel, consists of an inner sphere of fissile material, a layer of dense pyrolytic carbon (PyC), a ceramic layer of silicon carbide (SiC) and a final dense outer layer of PyC. These TRISO particles are then compacted with graphite into fuel rods that are then placed in channels in graphite blocks. The blocks are then arranged in an annular fashion to form a reactor core. However, this new fuel form has unanswered questions on the environmental post-burn-up behavior. The key question for current once-through fuel operations is how these large irradiated graphite blocks with spent fuel inside will behave in a repository environment. Data in the literature to answer this question is lacking, but nevertheless this is an important question that must be answered before wide-spread adoption of HTGR's could be considered. This research has focused on answering the question of how the large quantity of graphite surrounding the spent HTGR fuel will impact the release of aqueous uranium from the TRISO fuel. In order to answer this question, the sorption and partitioning behavior of uranium to graphite under a variety of conditions was investigated. Key systematic variables that were analyzed include solution pH, dissolved carbonate concentration, uranium metal concentration and ionic strength. The kinetics and desorption characteristics of uranium/graphite partitioning were studied as well. The graphite used in these experiments was also characterized by a variety of techniques and conclusions are drawn about the relevant surface chemistry of graphite. This data was then used to generate a model for the reactive transport of uranium in a graphite matrix. This model was implemented with the software code CXTFIT and validated through the use of column studies mirroring the predicted system.

A COOLED NEUTRONIC REACTOR

DOEpatents

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

1960-03-15

A nuclear reactor comprising a pair of graphite blocks separated by an air gap is described. Each of the blocks contains a plurality of channels extending from the gap through the block with a plurality of fuel elements being located in the channels. Means are provided for introducing air into the gap between the graphite blocks and for exhausting the air from the ends of the channels opposite the gap.

Monovacancy paramagnetism in neutron-irradiated graphite probed by ¹³C NMR.

PubMed

Zhang, Zhi Tao; Xu, C; Dmytriieva, Daryna; Molatta, Sebastian; Wosnitza, J; Wang, Y T; Helm, Manfred; Zhou, Shengqiang; Kuehne, Hannes

2017-09-18

We report on the magnetic properties of monovacancy defects in neutron-irradiated graphite, probed by $^{13}$C nuclear magnetic resonance spectroscopy. The bulk paramagnetism of the defect moments is revealed by the temperature dependence of the NMR frequency shift and spectral linewidth, both of which follow a Curie behavior, in agreement with measurements of the macroscopic magnetization. Compared to pristine graphite, the fluctuating hyperfine fields generated by the defect moments lead to an enhancement of the $^{13}$C nuclear spin-lattice relaxation rate $1/T_{1}$ by about two orders of magnitude. With an applied magnetic field of 7.1 T, the temperature dependence of $1/T_{1}$ below about 10 K can well be described by a thermally activated form, $1/T_{1}\\propto\\exp(-\\Delta/k_{B}T)$, yielding a singular Zeeman energy of ($0.41\\pm0.01$) meV, in excellent agreement with the sole presence of polarized, non-interacting defect moments. © 2017 IOP Publishing Ltd.

Compatibility of the Radio Frequency Mass Gauge with Composite Tanks

NASA Technical Reports Server (NTRS)

Zimmerli, Greg; Mueller, Carl

2015-01-01

The radio frequency mass gauge (RFMG) is a low-gravity propellant quantity gauge being developed at NASA for possible use in long-duration space missions utilizing cryogenic propellants. As part of the RFMG technology development process, we evaluated the compatibility of the RFMG with a graphite-epoxy composite material used to construct propellant tanks. The key material property that can affect compatibility with the RFMG is the electrical conductivity. Using samples of 8552IM7 graphite-epoxy composite, we characterized the resistivity and reflectivity over a range of frequencies. An RF impedance analyzer was used to characterize the out-of-plane electrical properties (along the sample thickness) in the frequency range 10 to 1800 MHZ. The resistivity value at 500 MHz was 4.8 ohm-cm. Microwave waveguide measurements of samples in the range 1.7 2.6 GHz, performed by inserting the samples into a WR-430 waveguide, showed reflectivity values above 98. Together, these results suggested that a tank constructed from graphite-epoxy composite would produce good quality electromagnetic tank modes, which is needed for the RFMG. This was verified by room-temperature measurements of the electromagnetic modes of a 2.4 m diameter tank constructed by Boeing from similar graphite-epoxy composite material. The quality factor Q of the tank electromagnetic modes, measured via RF reflection measurements from an antenna mounted in the tank, was typically in the range 400 Q 3000. The good quality modes observed in the tank indicate that the RFMG is compatible with graphite-epoxy tanks, and thus the RFMG could be used as a low-gravity propellant quantity gauge in such tanks filled with cryogenic propellants.

Mars Mission Analysis Trades Based on Legacy and Future Nuclear Propulsion Options

NASA Astrophysics Data System (ADS)

Joyner, Russell; Lentati, Andrea; Cichon, Jaclyn

2007-01-01

The purpose of this paper is to discuss the results of mission-based system trades when using a nuclear thermal propulsion (NTP) system for Solar System exploration. The results are based on comparing reactor designs that use a ceramic-metallic (CERMET), graphite matrix, graphite composite matrix, or carbide matrix fuel element designs. The composite graphite matrix and CERMET designs have been examined for providing power as well as propulsion. Approaches to the design of the NTP to be discussed will include an examination of graphite, composite, carbide, and CERMET core designs and the attributes of each in regards to performance and power generation capability. The focus is on NTP approaches based on tested fuel materials within a prismatic fuel form per the Argonne National Laboratory testing and the ROVER/NERVA program. NTP concepts have been examined for several years at Pratt & Whitney Rocketdyne for use as the primary propulsion for human missions beyond earth. Recently, an approach was taken to examine the design trades between specific NTP concepts; NERVA-based (UC)C-Graphite, (UC,ZrC)C-Composite, (U,Zr)C-Solid Carbide and UO2-W CERMET. Using Pratt & Whitney Rocketdyne's multidisciplinary design analysis capability, a detailed mission and vehicle model has been used to examine how several of these NTP designs impact a human Mars mission. Trends for the propulsion system mass as a function of power level (i.e. thrust size) for the graphite-carbide and CERMET designs were established and correlated against data created over the past forty years. These were used for the mission trade study. The resulting mission trades presented in this paper used a comprehensive modeling approach that captures the mission, vehicle subsystems, and NTP sizing.

Small Reactor Designs Suitable for Direct Nuclear Thermal Propulsion: Interim Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bruce G. Schnitzler

Advancement of U.S. scientific, security, and economic interests requires high performance propulsion systems to support missions beyond low Earth orbit. A robust space exploration program will include robotic outer planet and crewed missions to a variety of destinations including the moon, near Earth objects, and eventually Mars. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. In NASA's recent Mars Design Reference Architecture (DRA) 5.0 study, nuclear thermal propulsion (NTP) was again selectedmore » over chemical propulsion as the preferred in-space transportation system option for the human exploration of Mars because of its high thrust and high specific impulse ({approx}900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. The recently announced national space policy2 supports the development and use of space nuclear power systems where such systems safely enable or significantly enhance space exploration or operational capabilities. An extensive nuclear thermal rocket technology development effort was conducted under the Rover/NERVA, GE-710 and ANL nuclear rocket programs (1955-1973). Both graphite and refractory metal alloy fuel types were pursued. The primary and significantly larger Rover/NERVA program focused on graphite type fuels. Research, development, and testing of high temperature graphite fuels was conducted. Reactors and engines employing these fuels were designed, built, and ground tested. The GE-710 and ANL programs focused on an alternative ceramic-metallic 'cermet' fuel type consisting of UO2 (or UN) fuel embedded in a refractory metal matrix such as tungsten. The General Electric program examined closed loop concepts for space or terrestrial applications as well as open loop systems for direct nuclear thermal propulsion. Although a number of fast spectrum reactor and engine designs suitable for direct nuclear thermal propulsion were proposed and designed, none were built. This report summarizes status results of evaluations of small nuclear reactor designs suitable for direct nuclear thermal propulsion.« less

High-temperature compatibility study of iridium (DOP-26 alloy) with graphite and plutonia

DOE Office of Scientific and Technical Information (OSTI.GOV)

Axler, K.M.; Eash, D.T.

1987-12-01

This report outlines the materials compatibility tests conducted on DOP-26 iridium alloy and carbon. The carbon used was in the form of woven graphite as present in the impact shell used to encase plutonia in nuclear heat sources. In addition, compatibility tests of the DOP-26 alloy with plutonia are described. The reactivity observed in both systems is discussed. 4 refs., 6 figs.

Low temperature chemical processing of graphite-clad nuclear fuels

DOEpatents

Pierce, Robert A.

2017-10-17

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

Mineralogical and isotopic characterization of graphite deposits from the Anatectic Complex of Toledo, central Spain

NASA Astrophysics Data System (ADS)

Martín-Méndez, Iván; Boixereu, Ester; Villaseca, Carlos

2016-06-01

Graphite is found dispersed in high-grade metapelitic rocks of the Anatectic Complex of Toledo (ACT) and was mined during the mid twentieth century in places where it has been concentrated (Guadamur and la Puebla de Montalbán mines). Some samples from these mines show variable but significant alteration intensity, reaching very low-T hydrothermal (supergene) conditions for some samples from the waste heap of the Guadamur site (<100 °C and 1 kbar). Micro-Raman and XRD data indicate that all the studied ACT graphite is of high crystallinity irrespective of the degree of hydrothermal alteration. Chemical differences were obtained for graphite δ13C composition. ACT granulitic graphite shows δ13CPDB values in the range of -20.5 to -27.8 ‰, indicating a biogenic origin. Interaction of graphite with hydrothermal fluids does not modify isotopic compositions even in the most transformed samples from mining sites. The different isotopic signatures of graphite from the mining sites reflect its contrasted primary carbon source. The high crystallinity of studied graphite makes this area of central Spain suitable for graphitic exploration and its potential exploitation, due to the low carbon content required for its viability and its strategic applications in advanced technologies, such as graphene synthesis.

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

NASA Astrophysics Data System (ADS)

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

2016-08-01

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.

Modeling irradiation creep of graphite using rate theory

DOE PAGES

Sarkar, Apu; Eapen, Jacob; Raj, Anant; ...

2016-02-20

In this work we examined irradiation induced creep of graphite in the framework of transition state rate theory. Experimental data for two grades of nuclear graphite (H-337 and AGOT) were analyzed to determine the stress exponent (n) and activation energy (Q) for plastic flow under irradiation. Here we show that the mean activation energy lies between 0.14 and 0.32 eV with a mean stress-exponent of 1.0 ± 0.2. A stress exponent of unity and the unusually low activation energies strongly indicate a diffusive defect transport mechanism for neutron doses in the range of 3-4 x 10 22 n/cm 2.

Characterization of the Intrinsic Water Wettability of Graphite Using Contact Angle Measurements: Effect of Defects on Static and Dynamic Contact Angles.

PubMed

Kozbial, Andrew; Trouba, Charlie; Liu, Haitao; Li, Lei

2017-01-31

Elucidating the intrinsic water wettability of the graphitic surface has increasingly attracted research interests, triggered by the recent finding that the well-established hydrophobicity of graphitic surfaces actually results from airborne hydrocarbon contamination. Currently, static water contact angle (WCA) is often used to characterize the intrinsic water wettability of graphitic surfaces. In the current paper, we show that because of the existence of defects, static WCA does not necessarily characterize the intrinsic water wettability. Freshly exfoliated graphite of varying qualities, characterized using atomic force microscopy and Raman spectroscopy, was studied using static, advancing, and receding WCA measurements. The results showed that graphite of different qualities (i.e., defect density) always has a similar advancing WCA, but it could have very different static and receding WCAs. This finding indicates that defects play an important role in contact angle measurements, and the static contact angle does not always represent the intrinsic water wettability of pristine graphite. On the basis of the experimental results, a qualitative model is proposed to explain the effect of defects on static, advancing, and receding contact angles. The model suggests that the advancing WCA reflects the intrinsic water wettability of pristine (defect-free) graphite. Our results showed that the advancing WCA for pristine graphite is 68.6°, which indicates that graphitic carbon is intrinsically mildly hydrophilic.

Lunar sample analysis

NASA Technical Reports Server (NTRS)

Housley, R. M.

1983-01-01

The evolution of the lunar regolith under solar wind and micrometeorite bombardment is discussed as well as the size distribution of ultrafine iron in lunar soil. The most important characteristics of complex graphite, sulfide, arsenide, palladium, and platinum mineralization in a pegmatoid pyroxenite of the Stillwater Complex in Montana are examined. Oblique reflected light micrographs and backscattered electron SEM images of the graphite associations are included.

Temperature and flow fields in samples heated in monoellipsoidal mirror furnaces

NASA Astrophysics Data System (ADS)

Rivas, D.; Haya, R.

The temperature field in samples heated in monoellipsoidal mirror furnaces will be analyzed. The radiation heat exchange between the sample and the mirror is formulated analytically, taking into account multiple reflections at the mirror. It will be shown that the effect of these multiple reflections in the heating process is quite important, and, as a consequence, the effect of the mirror reflectance in the temperature field is quite strong. The conduction-radiation model will be used to simulate the heating process in the floating-zone technique in microgravity conditions; important parameters like the Marangoni number (that drives the thermocapillary flow in the melt), and the temperature gradient at the melt-crystal interface will be estimated. The model will be validated comparing with experimental data. The case of samples mounted in a wall-free configuration (as in the MAXUS-4 programme) will be also considered. Application to the case of compound samples (graphite-silicon-graphite) will be made; the melting of the silicon part and the surface temperature distribution in the melt will be analyzed. Of special interest is the temperature difference between the two graphite rods that hold the silicon part, since it drives the thermocapillary flow in the melt. This thermocapillary flow will be studied, after coupling the previous model with the convective effects. The possibility of counterbalancing this flow by the controlled vibration of the graphite rods will be studied as well. Numerical results show that suppressing the thermocapillary flow can be accomplished quite effectively.

Heat and mass transfer rates during flow of dissociated hydrogen gas over graphite surface

NASA Technical Reports Server (NTRS)

Nema, V. K.; Sharma, O. P.

1986-01-01

To improve upon the performance of chemical rockets, the nuclear reactor has been applied to a rocket propulsion system using hydrogen gas as working fluid and a graphite-composite forming a part of the structure. Under the boundary layer approximation, theoretical predictions of skin friction coefficient, surface heat transfer rate and surface regression rate have been made for laminar/turbulent dissociated hydrogen gas flowing over a flat graphite surface. The external stream is assumed to be frozen. The analysis is restricted to Mach numbers low enough to deal with the situation of only surface-reaction between hydrogen and graphite. Empirical correlations of displacement thickness, local skin friction coefficient, local Nusselt number and local non-dimensional heat transfer rate have been obtained. The magnitude of the surface regression rate is found low enough to ensure the use of graphite as a linear or a component of the system over an extended period without loss of performance.

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

PubMed Central

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

2018-01-01

14C is one of the limiting radionuclides used in the categorization of radioactive graphite waste; this categorization is crucial in selecting the appropriate graphite treatment/disposal method. We propose a rapid analysis method for 14C specific activity determination in small graphite samples in the 1–100 μg range. The method applies an oxidation procedure to the sample, which extracts 14C from the different carbonaceous matrices in a controlled manner. Because this method enables fast online measurement and 14C specific activity evaluation, it can be especially useful for characterizing 14C in irradiated graphite when dismantling graphite moderator and reflector parts, or when sorting radioactive graphite waste from decommissioned nuclear power plants. The proposed rapid method is based on graphite combustion and the subsequent measurement of both CO2 and 14C, using a commercial elemental analyser and the semiconductor detector, respectively. The method was verified using the liquid scintillation counting (LSC) technique. The uncertainty of this rapid method is within the acceptable range for radioactive waste characterization purposes. The 14C specific activity determination procedure proposed in this study takes approximately ten minutes, comparing favorably to the more complicated and time consuming LSC method. This method can be potentially used to radiologically characterize radioactive waste or used in biomedical applications when dealing with the specific activity determination of 14C in the sample. PMID:29370233

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

PubMed

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

2018-01-01

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

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

NASA Astrophysics Data System (ADS)

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

2018-04-01

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

Monitoring Prepregs As They Cure

NASA Technical Reports Server (NTRS)

Young, P. R.; Gleason, J. R.; Chang, A. C.

1986-01-01

Quality IR spectra obtained in dynamic heating environment. New technique obtains quality infrared spectra on graphite-fiber-reinforced, polymeric-matrix-resin prepregs as they cure. Technique resulted from modification of diffuse reflectance/Fourier transform infrared (DR/FTIR) technique previously used to analyze environmentally exposed cured graphite composites. Technique contribute to better understanding of prepreg chemistry/temperature relationships and development of more efficient processing cycles for advanced materials.

Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

2017-12-01

The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

Theory and application of laser ultrasonic shear wave birefringence measurements to the determination of microstructure orientation in transversely isotropic, polycrystalline graphite materials

DOE PAGES

Zeng, Fan W.; Contescu, Cristian I.; Gallego, Nidia C.; ...

2016-12-18

Laser ultrasonic line source methods have been used to study elastic anisotropy in nuclear graphites by measuring shear wave birefringence. Depending on the manufacturing processes used during production, nuclear graphites can exhibit various degrees of material anisotropy related to preferred crystallite orientation and to microcracking. In this paper, laser ultrasonic line source measurements of shear wave birefringence on NBG-25 have been performed to assess elastic anisotropy. Laser line sources allow specific polarizations for shear waves to be transmitted – the corresponding wavespeeds can be used to compute bulk, elastic moduli that serve to quantify anisotropy. These modulus values can bemore » interpreted using physical property models based on orientation distribution coefficients and microcrack-modified, single crystal moduli to represent the combined effects of crystallite orientation and microcracking on material anisotropy. Finally, ultrasonic results are compared to and contrasted with measurements of anisotropy based on the coefficient of thermal expansion to show the relationship of results from these techniques.« less

The effect of carbon crystal structure on treat reactor physics calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swanson, R.W.; Harrison, L.J.

1988-01-01

The Transient Reactor Test Facility (TREAT) at Argonne National Laboratory-West (ANL-W) is fueled with urania in a graphite and carbon mixture. This fuel was fabricated from a mixture of graphite flour, thermax (a thermatomic carbon produced by ''cracking'' natural gas), coal-tar resin and U/sub 3/O/sub 8/. During the fabrication process, the fuel was baked to dissociate the resin, but the high temperature necessary to graphitize the carbon in the thermax and in the resin was avoided. Therefore, the carbon crystal structure is a complex mixture of graphite particles in a nongraphitized elemental carbon matrix. Results of calculations using macroscopic carbonmore » cross sections obtained by mixing bound-kernel graphite cross sections for the graphitized carbon and free-gas carbon cross sections for the remainder of the carbon and calculations using only bound-kernel graphite cross sections are compared to experimental data. It is shown that the use of the hybridized cross sections which reflect the allotropic mixture of the carbon in the TREAT fuel results in a significant improvement in the accuracy of calculated neutronics parameters for the TREAT reactor. 6 refs., 2 figs., 3 tabs.« less

NUCLEAR REACTOR

DOEpatents

Starr, C.

1963-01-01

This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)

Secondary electron emission characteristics of ion-textured copper and high-purity isotropic graphite surfaces

NASA Technical Reports Server (NTRS)

Curren, A. N.; Jensen, K. A.

1984-01-01

Experimentally determined values of true secondary electron emission and relative values of reflected primary electron yield for untreated and ion textured oxygen free high conductivity copper and untreated and ion textured high purity isotropic graphite surfaces are presented for a range of primary electron beam energies and beam impingement angles. This investigation was conducted to provide information that would improve the efficiency of multistage depressed collectors (MDC's) for microwave amplifier traveling wave tubes in space communications and aircraft applications. For high efficiency, MDC electrode surfaces must have low secondary electron emission characteristics. Although copper is a commonly used material for MDC electrodes, it exhibits relatively high levels of secondary electron emission if its surface is not treated for emission control. Recent studies demonstrated that high purity isotropic graphite is a promising material for MDC electrodes, particularly with ion textured surfaces. The materials were tested at primary electron beam energies of 200 to 2000 eV and at direct (0 deg) to near grazing (85 deg) beam impingement angles. True secondary electron emission and relative reflected primary electron yield characteristics of the ion textured surfaces were compared with each other and with those of untreated surfaces of the same materials. Both the untreated and ion textured graphite surfaces and the ion treated copper surface exhibited sharply reduced secondary electron emission characteristics relative to those of untreated copper. The ion treated graphite surface yielded the lowest emission levels.

Pile construction

DOEpatents

Johnson, Alfred A.; Carleton, John T.

1978-05-02

A graphite-moderated, water-cooled nuclear reactor including graphite blocks disposed in transverse alternate layers, one set of alternate layers consisting of alternate full size blocks and smaller blocks through which cooling tubes containing fuel extend, said smaller blocks consisting alternately of tube bearing blocks and support block, the support blocks being smaller than the tube bearing blocks, the aperture of each support block being tapered so as to provide the tube extending therethrough with a narrow region of support while being elsewhere spaced therefrom.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Loyalka, Sudarshan

High and Very High Temperatures Gas Reactors (HTGRs/VHTRs) have five barriers to fission product (FP) release: the TRISO fuel coating, the fuel elements, the core graphite, the primary coolant system, and the reactor building. This project focused on measurements and computations of FP diffusion in graphite, FP adsorption on graphite and FP interactions with dust particles of arbitrary shape. Diffusion Coefficients of Cs and Iodine in two nuclear graphite were obtained by the release method and use of Inductively Coupled Plasma-Mass Spectroscopy (ICP-MS) and Instrumented Neutron Activation Analysis (INAA). A new mathematical model for fission gas release from nuclear fuelmore » was also developed. Several techniques were explored to measure adsorption isotherms, notably a Knudsen Effusion Mass Spectrometer (KEMS) and Instrumented Neutron Activation Analysis (INAA). Some of these measurements are still in progress. The results will be reported in a supplemental report later. Studies of FP interactions with dust and shape factors for both chain-like particles and agglomerates over a wide size range were obtained through solutions of the diffusion and transport equations. The Green's Function Method for diffusion and Monte Carlo technique for transport were used, and it was found that the shape factors are sensitive to the particle arrangements, and that diffusion and transport of FPs can be hindered. Several journal articles relating to the above work have been published, and more are in submission and preparation.« less

Irradiation Creep in Graphite

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 similarlymore » 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.« less

NEUTRONIC REACTORS

DOEpatents

Wigner, E.P.

1960-11-22

A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

Treatment of irradiated graphite from French Bugey reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stevens, Howard; Laurent, Gerard

In 2008, following the general French plan for nuclear waste management, Electricite de France attempted to find for irradiated graphite an alternative solution to direct storage at the low-activity long-life storage center in France managed by the national agency for wastes (ANDRA). EDF management requested that its engineering arm, EDF CIDEN, study the graphite treatment alternatives to direct storage. In mid-2008, this study revealed the potential advantage for EDF to use a steam reforming process known as Thermal Organic Reduction, 'THOR' (owned by Studsvik, Inc., USA), to treat or destroy the graphite matrix and limit the quantity of secondary wastemore » to be stored. In late 2009, EDF began a test program with Studsvik to determine if the THOR steam reforming process could be used to destroy the graphite. The program also sought to determine if the graphite could be treated to release the bulk of activity while minimizing the gasification of the bulk mass of the graphite. In October 2009, tests with non-irradiated graphite were completed and demonstrated destruction of a graphite matrix by the THOR process at satisfactory rates. After gasifying the graphite, focus shifted to the effect of roasting graphite at high temperatures in inert gases with low concentrations of oxidizing gases to preferentially remove volatile radionuclides while minimizing the graphite mass loss to 5%. A radioactive graphite sleeve was imported from France to the US for these tests. Completed in April 2010, 'Phase I' of testing showed that the process removed >99% of H-3 and 46% of C-14 with <6% mass loss. Completed in September 2011, 'Phase II' testing achieved increased removals as high as 80% C-14. During Phase II, it was also discovered that roasting in a reducing atmosphere helped to limit the oxidation of the graphite. Future work seeks to explore the effects of reducing gases to limit the bulk oxidation of graphite. If the graphite could be decontaminated of long-lived radionuclides up to 95% for C-14 while minimizing mass loss to <5%, this would minimize the volume of any secondary waste streams and potentially lower the waste class of the larger bulk of graphite. Alternatively, if up to 95% decontamination of C-14 is achieved, the graphite may be completely gasified which could result in lower disposal. (authors)« less

GAS COOLED NUCLEAR REACTORS

DOEpatents

Long, E.; Rodwell, W.

1958-06-10

A gas-cooled nuclear reactor consisting of a graphite reacting core and reflector structure supported in a containing vessel is described. A gas sealing means is included for sealing between the walls of the graphite structure and containing vessel to prevent the gas coolant by-passing the reacting core. The reacting core is a multi-sided right prismatic structure having a pair of parallel slots around its periphery. The containing vessel is cylindrical and has a rib on its internal surface which supports two continuous ring shaped flexible web members with their radially innermost ends in sealing engagement within the radially outermost portion of the slots. The core structure is supported on ball bearings. This design permits thermal expansion of the core stracture and vessel while maintainirg a peripheral seal between the tvo elements.

Physicochemical characterization, and relaxometry studies of micro-graphite oxide, graphene nanoplatelets, and nanoribbons.

PubMed

Paratala, Bhavna S; Jacobson, Barry D; Kanakia, Shruti; Francis, Leonard Deepak; Sitharaman, Balaji

2012-01-01

The chemistry of high-performance magnetic resonance imaging contrast agents remains an active area of research. In this work, we demonstrate that the potassium permanganate-based oxidative chemical procedures used to synthesize graphite oxide or graphene nanoparticles leads to the confinement (intercalation) of trace amounts of Mn(2+) ions between the graphene sheets, and that these manganese intercalated graphitic and graphene structures show disparate structural, chemical and magnetic properties, and high relaxivity (up to 2 order) and distinctly different nuclear magnetic resonance dispersion profiles compared to paramagnetic chelate compounds. The results taken together with other published reports on confinement of paramagnetic metal ions within single-walled carbon nanotubes (a rolled up graphene sheet) show that confinement (encapsulation or intercalation) of paramagnetic metal ions within graphene sheets, and not the size, shape or architecture of the graphitic carbon particles is the key determinant for increasing relaxivity, and thus, identifies nano confinement of paramagnetic ions as novel general strategy to develop paramagnetic metal-ion graphitic-carbon complexes as high relaxivity MRI contrast agents.

Physicochemical Characterization, and Relaxometry Studies of Micro-Graphite Oxide, Graphene Nanoplatelets, and Nanoribbons

PubMed Central

Paratala, Bhavna S.; Jacobson, Barry D.; Kanakia, Shruti; Francis, Leonard Deepak; Sitharaman, Balaji

2012-01-01

The chemistry of high-performance magnetic resonance imaging contrast agents remains an active area of research. In this work, we demonstrate that the potassium permanganate-based oxidative chemical procedures used to synthesize graphite oxide or graphene nanoparticles leads to the confinement (intercalation) of trace amounts of Mn2+ ions between the graphene sheets, and that these manganese intercalated graphitic and graphene structures show disparate structural, chemical and magnetic properties, and high relaxivity (up to 2 order) and distinctly different nuclear magnetic resonance dispersion profiles compared to paramagnetic chelate compounds. The results taken together with other published reports on confinement of paramagnetic metal ions within single-walled carbon nanotubes (a rolled up graphene sheet) show that confinement (encapsulation or intercalation) of paramagnetic metal ions within graphene sheets, and not the size, shape or architecture of the graphitic carbon particles is the key determinant for increasing relaxivity, and thus, identifies nano confinement of paramagnetic ions as novel general strategy to develop paramagnetic metal-ion graphitic-carbon complexes as high relaxivity MRI contrast agents. PMID:22685555

A Comparison of the Irradiation Creep Behavior of Several Graphites

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burchell, Timothy D; Windes, Will

2016-01-01

Graphite creep strain data from the irradiation creep capsule Advanced Graphite Creep-1 (AGC-1) are reported. This capsule was the first (prototype) of a series of five or six capsules planned as part of the AGC experiment, which was designed to fully characterize the effects of neutron irradiation and the radiation creep behavior of current nuclear graphite. The creep strain data and analysis are reported for the six graphite grades incorporated in the capsule. The AGC-1 capsule was irradiated in the Advanced Test Reactor at Idaho National Laboratory (INL) at approximately 700 C and to a peak dose of 7 dpamore » (displacements per atom). The specimen s final dose, temperature, and stress conditions have been reported by INL and were used during this analysis. The derived creep coefficients (K) were calculated for each grade and were found to compare well to literature data for the creep coefficient, even under the wide range of AGC-1 specimen temperatures. Comparisons were made between AGC-1 data and historical grade data for creep coefficients.« less

Reversible Intercalation of Fluoride-Anion Receptor Complexes in Graphite

NASA Technical Reports Server (NTRS)

West, William C.; Whitacre, Jay F.; Leifer, Nicole; Greenbaum, Steve; Smart, Marshall; Bugga, Ratnakumar; Blanco, Mario; Narayanan, S. R.

2007-01-01

We have demonstrated a route to reversibly intercalate fluoride-anion receptor complexes in graphite via a nonaqueous electrochemical process. This approach may find application for a rechargeable lithium-fluoride dual-ion intercalating battery with high specific energy. The cell chemistry presented here uses graphite cathodes with LiF dissolved in a nonaqueous solvent through the aid of anion receptors. Cells have been demonstrated with reversible cathode specific capacity of approximately 80 mAh/g at discharge plateaus of upward of 4.8 V, with graphite staging of the intercalant observed via in situ synchrotron X-ray diffraction during charging. Electrochemical impedance spectroscopy and B-11 nuclear magnetic resonance studies suggest that cointercalation of the anion receptor with the fluoride occurs during charging, which likely limits the cathode specific capacity. The anion receptor type dictates the extent of graphite fluorination, and must be further optimized to realize high theoretical fluorination levels. To find these optimal anion receptors, we have designed an ab initio calculations-based scheme aimed at identifying receptors with favorable fluoride binding and release properties.

Nuclear fuel element

DOEpatents

Meadowcroft, Ronald Ross; Bain, Alastair Stewart

1977-01-01

A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops. A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber. The spacer is of zirconium or a zirconium alloy.

Thermal neutron streaming effects and WIMS analysis of the Penn State subcritical graphite pile

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Zediak, C.S.; Jester, W.A.

1997-12-01

This analysis was performed on the Pennsylvania State University (PSU) subcritical reactor to find more accurate values for such nuclear parameters as the thermal fuel utilization factor, thermal diffusion length in the graphite, migration area, k{sub eff}, etc. The analysis involved using the Winfrith Integrated Multigroup Scheme (WIMS) code as well as various hand calculations to find and compare those parameters. The data found in this analysis will be used by future students in the Penn State laboratory courses.

THE HOT CRITICAL ASSEMBLY $sub 4$CESAR$sub 4$ (in French)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tanguy, P.

1963-07-01

With Cesar, the Cadarache Center for Nuclear Studies will be equipped with a zero-power critical assembly, which will enable it to obtain the data necessary for the development of natural uranium, graphite, gas reactors. Reactivity balance, evolution of the reactivity, and deformation of the flux curves are to be studied. These studies will complement those already being done on Marius, but carried out at room temperature; in Cesar the graphite temperature can reach 500 deg C. (auth)

The Nuclear Cryogenic Propulsion Stage

NASA Technical Reports Server (NTRS)

Houts, Michael G.; Kim, Tony; Emrich, William J.; Hickman, Robert R.; Broadway, Jeramie W.; Gerrish, Harold P.; Belvin, Anthony D.; Borowski, Stanley K.; Scott, John H.

2014-01-01

Nuclear Thermal Propulsion (NTP) development efforts in the United States have demonstrated the technical viability and performance potential of NTP systems. For example, Project Rover (1955 - 1973) completed 22 high power rocket reactor tests. Peak performances included operating at an average hydrogen exhaust temperature of 2550 K and a peak fuel power density of 5200 MW/m3 (Pewee test), operating at a thrust of 930 kN (Phoebus-2A test), and operating for 62.7 minutes in a single burn (NRX-A6 test). Results from Project Rover indicated that an NTP system with a high thrust-to-weight ratio and a specific impulse greater than 900 s would be feasible. Excellent results were also obtained by the former Soviet Union. Although historical programs had promising results, many factors would affect the development of a 21st century nuclear thermal rocket (NTR). Test facilities built in the US during Project Rover no longer exist. However, advances in analytical techniques, the ability to utilize or adapt existing facilities and infrastructure, and the ability to develop a limited number of new test facilities may enable affordable development, qualification, and utilization of a Nuclear Cryogenic Propulsion Stage (NCPS). Bead-loaded graphite fuel was utilized throughout the Rover/NERVA program, and coated graphite composite fuel (tested in the Nuclear Furnace) and cermet fuel both show potential for even higher performance than that demonstrated in the Rover/NERVA engine tests.. NASA's NCPS project was initiated in October, 2011, with the goal of assessing the affordability and viability of an NCPS. FY 2014 activities are focused on fabrication and test (non-nuclear) of both coated graphite composite fuel elements and cermet fuel elements. Additional activities include developing a pre-conceptual design of the NCPS stage and evaluating affordable strategies for NCPS development, qualification, and utilization. NCPS stage designs are focused on supporting human Mars missions. The NCPS is being designed to readily integrate with the Space Launch System (SLS). A wide range of strategies for enabling affordable NCPS development, qualification, and utilization should be considered. These include multiple test and demonstration strategies (both ground and in-space), multiple potential test sites, and multiple engine designs. Two potential NCPS fuels are currently under consideration - coated graphite composite fuel and tungsten cermet fuel. During 2014 a representative, partial length (approximately 16") coated graphite composite fuel element with prototypic depleted uranium loading is being fabricated at Oak Ridge National Laboratory (ORNL). In addition, a representative, partial length (approximately 16") cermet fuel element with prototypic depleted uranium loading is being fabricated at Marshall Space Flight Center (MSFC). During the development process small samples (approximately 3" length) will be tested in the Compact Fuel Element Environmental Tester (CFEET) at high temperature (approximately 2800 K) in a hydrogen environment to help ensure that basic fuel design and manufacturing process are adequate and have been performed correctly. Once designs and processes have been developed, longer fuel element segments will be fabricated and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREE) at high temperature (approximately 2800 K) and in flowing hydrogen.

10 CFR 110.70 - Public notice of receipt of an application.

Code of Federal Regulations, 2010 CFR

2010-01-01

... MATERIAL Public Notification and Availability of Documents and Records § 110.70 Public notice of receipt of... kilograms or more of heavy water. (4) Nuclear grade graphite for nuclear end use. (5) Radioactive waste. (Note: Does not apply to exports of heavy water to Canada.) (c) The Commission will also publish in the...

Notched graphite polymimide composites at room and notched graphite polymide composites at room and elevated temperatures. [nondestructive test techniques

NASA Technical Reports Server (NTRS)

Awerbuch, J.; Perkinson, H. E.; Kamel, I. L.

1980-01-01

The fracture behavior in graphite/polyimide (Gr/PI) Celion 6000/PMR-15 composites was characterized. Emphasis was placed on the correlation between the observed failure modes and the deformation characteristics of center-notched Gr/Pl laminates. Crack tip damage growth, fracture strength and notch sensitivity, and the associated characterization methods were also examined. Special attention was given to nondestructive evaluation of internal damage and damage growth, techniques such as acoustic emission, X-ray radiography, and ultrasonic C-scan. Microstructural studies using scanning electron microscopy, photomicrography, and the pulsed nuclear magnetic resonance technique were employed as well. All experimental procedures and techniques are described and a summary of representative results for Gr/Pl laminates is given.

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

NASA Astrophysics Data System (ADS)

McKenna, Alice

One of the functions of graphite is as a moderator in several nuclear reactor designs, including the Advanced Gas-cooled Reactor (AGR). In the reactor graphite is used to thermalise the neutrons produced in the fission reaction thus allowing a self-sustained reaction to occur. The graphite blocks, acting as the moderator, are constantly irradiated and consequently suffer damage. This thesis examines the types of damage caused using molecular dynamic (MD) simulations and ab intio calculations. Neutron damage starts with a primary knock-on atom (PKA), which is travelling so fast that it creates damage through electronic and thermal excitation (this is addressed with thermal spike simulations). When the PKA has lost energy the subsequent cascade is based on ballistic atomic displacement. These two types of simulations were performed on single crystal graphite and other carbon structures such as diamond and amorphous carbon as a comparison. The thermal spike in single crystal graphite produced results which varied from no defects to a small number of permanent defects in the structure. It is only at the high energy range that more damage is seen but these energies are less likely to occur in the nuclear reactor. The thermal spike does not create damage but it is possible that it can heal damaged sections of the graphite, which can be demonstrated with the motion of the defects when a thermal spike is applied. The cascade simulations create more damage than the thermal spike even though less energy is applied to the system. A new damage function is found with a threshold region that varies with the square root of energy in excess of the energy threshold. This is further broken down in to contributions from primary and subsequent knock-on atoms. The threshold displacement energy (TDE) is found to be Ed=25eV at 300K. In both these types of simulation graphite acts very differently to the other carbon structures. There are two types of polycrystalline graphite structures which simulations have been performed on. The difference between the two is at the grain boundaries with one having dangling bonds and the other one being bonded. The cascade showed the grain boundaries acting as a trap for the knock-on atoms which produces more damage compared with the single crystal. Finally the effects of turbostratic disorder on damage is considered. Density functional theory (DFT) was used to look at interstitials in (002) twist boundaries and how they act compared to AB stacked graphite. The results of these calculations show that the spiro interstitial is more stable in these grain boundaries, so at temperatures where the interstitial can migrate along the c direction they will segregate to (002) twist boundaries.

Nuclear Graphite - Fracture Behavior and Modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burchell, Timothy D; Battiste, Rick; Strizak, Joe P

2011-01-01

Evidence for the graphite fracture mechanism is reviewed and discussed. The roles of certain microstructural features in the graphite fracture process are reported. The Burchell fracture model is described and its derivation reported. The successful application of the fracture model to uniaxial tensile data from several graphites with widely ranging structure and texture is reported. The extension of the model to multiaxial loading scenarios using two criteria is discussed. Initially, multiaxial strength data for H-451 graphite were modeled using the fracture model and the Principle of Independent Action. The predicted 4th stress quadrant failure envelope was satisfactory but the 1stmore » quadrant predictions were not conservative and thus were unsatisfactory. Multiaxial strength data from the 1st and 4th stress quadrant for NBG-18 graphite are reported. To improve the conservatism of the predicted 1st quadrant failure envelope for NBG-18 the Shetty criterion has been applied to obtain the equivalent critical stress intensity factor, KIc (Equi), for each applied biaxial stress ratio. The equivalent KIc value is used in the Burchell fracture model to predict the failure envelope. The predicted 1st stress quadrant failure envelope is conservative and thus more satisfactory than achieved previously using the fracture model combined with the Principle of Independent Action.« less

Statistical Models of Fracture Relevant to Nuclear-Grade Graphite: Review and Recommendations

NASA Technical Reports Server (NTRS)

Nemeth, Noel N.; Bratton, Robert L.

2011-01-01

The nuclear-grade (low-impurity) graphite needed for the fuel element and moderator material for next-generation (Gen IV) reactors displays large scatter in strength and a nonlinear stress-strain response from damage accumulation. This response can be characterized as quasi-brittle. In this expanded review, relevant statistical failure models for various brittle and quasi-brittle material systems are discussed with regard to strength distribution, size effect, multiaxial strength, and damage accumulation. This includes descriptions of the Weibull, Batdorf, and Burchell models as well as models that describe the strength response of composite materials, which involves distributed damage. Results from lattice simulations are included for a physics-based description of material breakdown. Consideration is given to the predicted transition between brittle and quasi-brittle damage behavior versus the density of damage (level of disorder) within the material system. The literature indicates that weakest-link-based failure modeling approaches appear to be reasonably robust in that they can be applied to materials that display distributed damage, provided that the level of disorder in the material is not too large. The Weibull distribution is argued to be the most appropriate statistical distribution to model the stochastic-strength response of graphite.

Characterization of Hypervelocity Impact Debris from the DebriSat Tests

NASA Astrophysics Data System (ADS)

Adams, P. M.; Sheaffer, P. M.; Lingley, Z.; Radhakrishnan, G.

The DebriSat program consisted of 3 hypervelocity impact tests conducted in 2 Torr of air with 7 km/s, 600 g aluminum projectiles. In the first test, Pre Preshot, the target consisted of multiple layers of fiberglass, stainless steel and Kevlar fabric. No soft catch foam was used. The subsequent two tests, DebrisLV and DebriSat, were designed to simulate hypervelocity impacts with a launch vehicle upper stage and a modern LEO satellite, respectively. The interior of the chamber was lined with soft catch foam to trap break-up fragments. In all three tests, witness plates were placed near the target to sample impact debris and determine its reflectance, composition and spectral properties. Reflectance measurements are important for calculating the size of orbital hypervelocity impact fragments. The debris from the Pre Preshot test consisted of a two-phase mixture formed from solidified molten silicate and steel droplets. Individual droplets ranged from 100 μm to 10 nm. The reflectance of witness plates dropped from 95% to 20-30% as a result of the debris. Debris collected on witness plates in the DebrisLV and DebriSat tests consisted of μm to nm-sized solidified molten metallic droplets in a matrix of condensed vaporized soft catch. Disordered graphitic carbon was also detected. The reflectance of debris-covered witness plates dropped from 95% to 5%. The dramatic decrease in reflectance for hypervelocity impact debris is attributed to the effect of scattering from μm to nm sized solidified molten metallic droplets and the presence of graphitic carbon, when organics are present. The presence of soft catch in the later tests and the high organic content with graphitic carbon in the debris appear to be responsible for this much lower post-test reflectance. Understanding orbital debris reflectance is critical for estimating size and determining debris detectability.

Nuclear Cryogenic Propulsion Stage (NCPS) Fuel Element Testing in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2017-01-01

To satisfy the Nuclear Cryogenic Propulsion Stage (NCPS) testing milestone, a graphite composite fuel element using a uranium simulant was received from the Oakridge National Lab and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) at various operating conditions. The nominal operating conditions required to satisfy the milestone consisted of running the fuel element for a few minutes at a temperature of at least 2000 K with flowing hydrogen. This milestone test was successfully accomplished without incident.

Ukraine’s Defense Dilemma

DTIC Science & Technology

1992-08-25

Ukraine has made environmental issues a top priority in any national security equation. The Chernobyl disaster of April 1986 is estimated to have...investment in Ukraine and in other former Soviet republics may be jeopardized by another environmental disaster of Chernobyl proportions. As an energy...adequate nuclear plants nor ailternative energy sources beyond coal to substitute for nuclear generated electricity. Some 16 Chernobyl -type graphite

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

PubMed

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

2016-01-01

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).

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

PubMed Central

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

2016-01-01

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). PMID:27706228

Absorber for microwave investigation in the open space

NASA Astrophysics Data System (ADS)

Kubacki, Roman; Smólski, Bogusław; Głuszewski, Wojciech; Przesmycki, Rafał; Rudyk, Karol

2017-04-01

In some circumstances there is a need to realize the Electromagnetic Compatibility (EMC) investigation not in the specialized anechoic chamber but in the open space. Typical absorbers used in anechoic chamber to reduce the reflected rays from walls and floor, such as ferrite plates and graphite cones, are not suitable in the open space. In the work the investigation of the flexible absorbing material intended to the liquidation of the radiation reflected from the ground has been presented. As an absorbing material the metallic-glass with graphite was elaborated. This material was additionally exposed to the ionizing radiation in the dose of 100 kGy in the radioactive gamma source. The permittivity, permeability as well as the shielding properties have been analyzed.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Campbell, Anne A.; Katoh, Yutai; Snead, Mary A.

A new, fine-grain nuclear graphite, grade G347A from Tokai Carbon Co., Ltd., has been irradiated in the High Flux Isotope Reactor at Oak Ridge National Laboratory to study the materials property changes that occur when exposed to neutron irradiation at temperatures of interest for Generation-IV nuclear reactor applications. Specimen temperatures ranged from 290°C to 800 °C with a maximum neutron fluence of 40 × 10 25 n/m 2 [E > 0.1 MeV] (~30dpa). Lastly, observed behaviors include: anisotropic behavior of dimensional change in an isotropic graphite, Young's modulus showing parabolic fluence dependence, electrical resistivity increasing at low fluence and additionalmore » increase at high fluence, thermal conductivity rapidly decreasing at low fluence followed by continued degradation, and a similar plateau value of the mean coefficient of thermal expansion for all irradiation temperatures.« less

Modeling of irradiated graphite (14)C transfer through engineered barriers of a generic geological repository in crystalline rocks.

PubMed

Poskas, Povilas; Grigaliuniene, Dalia; Narkuniene, Asta; Kilda, Raimondas; Justinavicius, Darius

2016-11-01

There are two RBMK-1500 type graphite moderated reactors at the Ignalina nuclear power plant in Lithuania, and they are under decommissioning now. The graphite cannot be disposed of in a near surface repository, because of large amounts of (14)C. Therefore, disposal of the graphite in a geological repository is a reasonable solution. This study presents evaluation of the (14)C transfer by the groundwater pathway into the geosphere from the irradiated graphite in a generic geological repository in crystalline rocks and demonstration of the role of the different components of the engineered barrier system by performing local sensitivity analysis. The speciation of the released (14)C into organic and inorganic compounds as well as the most recent information on (14)C source term was taken into account. Two alternatives were considered in the analysis: disposal of graphite in containers with encapsulant and without it. It was evaluated that the maximal fractional flux of inorganic (14)C into the geosphere can vary from 10(-11)y(-1) (for non-encapsulated graphite) to 10(-12)y(-1) (for encapsulated graphite) while of organic (14)C it was about 10(-3)y(-1) of its inventory. Such difference demonstrates that investigations on the (14)C inventory and chemical form in which it is released are especially important. The parameter with the highest influence on the maximal flux into the geosphere for inorganic (14)C transfer was the sorption coefficient in the backfill and for organic (14)C transfer - the backfill hydraulic conductivity. Copyright © 2016 Elsevier B.V. All rights reserved.

Graphene nanosheets and graphite oxide as promising adsorbents for removal of organic contaminants from aqueous solution.

PubMed

Ji, Liangliang; Chen, Wei; Xu, Zhaoyi; Zheng, Shourong; Zhu, Dongqiang

2013-01-01

Graphenes are an emerging class of carbon nanomaterials whose adsorption properties toward organic compounds have not been well understood. In the present study, graphene nanosheets were prepared by reoxidation and abrupt heating of graphite oxide, which was prepared by sequential chemical oxidation of commercial nonporous graphite powder. Adsorption properties of three aromatic compounds (naphthalene, 2-naphthol, and 1-naphthylamine) and one pharmaceutical compound (tylosin) on graphene nanosheets and graphite oxide were examined to explore the potential of these two adsorbents for the removal of organic contaminants from aqueous solutions. Compared with the literature data of adsorption on carbon nanotubes, adsorption of bulky, flexible tylosin on graphene nanosheets exhibited markedly faster adsorption kinetics, which can be attributed to their opened-up layer structure. Graphene nanosheets and graphite oxide showed similar sequences of adsorption affinity: 1-naphthylamine > 2-naphthol > tylosin > naphthalene (with much larger differences observed on graphite oxide). It was proposed that the strong adsorption of the three aromatic compounds was mainly due to π-π electron donor-acceptor interactions with the graphitic surfaces of adsorbents. Additionally, Lewis acid-base interaction was likely an important factor contributing to the strong adsorption of 1-naphthylamine and tylosin, especially for the O-functionality-abundant graphite oxide. After being normalized on the basis of adsorbent surface area, adsorption affinities of all four tested adsorbates on graphene nanosheets were very close to those on nonporous graphite powder, reflecting complete accessibility of the adsorbent surface area in adsorption. Copyright © by the American Society of Agronomy, Crop Science Society of America, and Soil Science Society of America, Inc.

Basic experiments during loss of vacuum event (LOVE) in fusion experimental reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ogawa, Masuro; Kunugi, Tomoaki; Seki, Yasushi

If a loss of vacuum event (LOVE) occurs due to damage of the vacuum vessel of a nuclear fusion experimental reactor, some chemical reactions such as a graphic oxidation and a buoyancy-driven exchange flow take place after equalization of the gas pressure between the inside and outside of the vacuum vessel. The graphite oxidation would generate inflammable carbon monoxide and release tritium retained in the graphite. The exchange flow through the breaches may transport the carbon monoxide and tritium out of the vacuum vessel. To add confidence to the safety evaluations and analyses, it is important to grasp the basicmore » phenomena such as the exchange flow and the graphite oxidation. Experiments of the exchange flow and the graphite oxidation were carried out to obtain the exchange flow rate and the rate constant for the carbon monoxide combustion, respectively. These experimental results were compared with existing correlations. The authors plan a scaled-model test and a full-scale model test for the LOVE.« less

Chemical Characterization and Removal of C-14 from Irradiated Graphite-12010

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cleaver, James; McCrory, Shilo; Smith, Tara E.

2012-07-01

Quantities of irradiated graphite waste are expected to drastically increase, which indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 (C-14), with a half-life of 5730 years. Study of irradiated graphite from nuclear reactors indicates C-14 is concentrated on the outer 5 mm of the graphite structure. The aim of the research described here is to identify the chemical form of C-14 in irradiated graphite and develop a practical method by which C-14 can be removed. Characterization of pre- and post-irradiation graphite was conducted to determine bond type, functionalmore » groups, location and concentration of C-14 and its precursors via the use of surface sensitive characterization techniques. Because most surface C-14 originates from neutron activation of nitrogen, an understanding of nitrogen bonding to graphite may lead to a greater understanding of the formation pathway of C-14. However, no single technique provides a complete picture. Therefore, a portfolio of techniques has been developed, with each technique providing another piece to the puzzle that is the chemical nature of the C-14. Scanning Electron Microscopy (SEM), X-Ray Diffraction (XRD), and Raman Spectroscopy were used to evaluate the morphological features of graphite samples. The concentration, chemical composition, and bonding characteristics of C-14 and its precursors were determined through X-ray Photoelectron Spectroscopy (XPS), Time-of-Flight Secondary Ion Mass Spectrometry (SIMS), and Auger and Energy Dispersive X-ray Analysis Spectroscopy (EDX). High-surface-area graphite foam, POCOFoam{sup R}, was exposed to liquid nitrogen and irradiated. Characterization of this material has shown C-14 to C-12 ratios of 0.035. This information was used to optimize the thermal treatment of graphite. Thermal treatment of irradiated graphite as reported by Fachinger et al. (2007) uses naturally adsorbed oxygen complexes to gasify graphite, thus its effectiveness is highly dependent on the availability of adsorbed oxygen compounds. In research presented, the quantity and form of adsorbed oxygen complexes in pre- and post irradiated graphite was studied using SIMS and XPS. SIMS and XPS detected adsorbed oxygen compounds on both irradiated and unirradiated graphite. During thermal treatment graphite samples are heated in the presence of inert argon gas, which carries off gaseous products released during treatment. Experiments were performed at 900 deg. C and 1400 deg. C to evaluate the selective removal of C-14. (authors)« less

EXPERIMENTAL LIQUID METAL FUEL REACTOR

DOEpatents

Happell, J.J.; Thomas, G.R.; Denise, R.P.; Bunts, J.L. Jr.

1962-01-23

A liquid metal fuel nuclear fission reactor is designed in which the fissionable material is dissolved or suspended in a liquid metal moderator and coolant. The liquid suspension flows into a chamber in which a critical amount of fissionable material is obtained. The fluid leaves the chamber and the heat of fission is extracted for power or other utilization. The improvement is in the support arrangement for a segrnented graphite core to permit dif ferential thermal expansion, effective sealing between main and blanket liquid metal flows, and avoidance of excessive stress development in the graphite segments. (AEC)

Coupled field-structural analysis of HGTR fuel brick using ABAQUS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohanty, S.; Jain, R.; Majumdar, S.

2012-07-01

High-temperature, gas-cooled reactors (HTGRs) are usually helium-gas cooled, with a graphite core that can operate at reactor outlet temperatures much higher than can conventional light water reactors. In HTGRs, graphite components moderate and reflect neutrons. During reactor operation, high temperature and high irradiation cause damage to the graphite crystal and grains and create other defects. This cumulative structural damage during the reactor lifetime leads to changes in graphite properties, which can alter the ability to support the designed loads. The aim of the present research is to develop a finite-element code using commercially available ABAQUS software for the structural integritymore » analysis of graphite core components under extreme temperature and irradiation conditions. In addition, the Reactor Geometry Generator tool-kit, developed at Argonne National Laboratory, is used to generate finite-element mesh for complex geometries such as fuel bricks with multiple pin holes and coolant flow channels. This paper presents the proposed concept and discusses results of stress analysis simulations of a fuel block with H-451 grade material properties. (authors)« less

Forgery at the Snite Museum of Art? Improving AMS Radiocarbon Dating at the University of Notre Dame

NASA Astrophysics Data System (ADS)

Troyer, Laura; Bagwell, Connor; Anderson, Tyler; Clark, Adam; Nelson, Austin; Skulski, Michael; Collon, Philippe

2017-09-01

The Snite Museum of Art recently obtained several donations of artifacts. Five of the pieces lack sufficient background information to prove authenticity and require further analysis to positively determine the artwork's age. One method to determine the artwork's age is radiocarbon dating via Accelerator Mass Spectrometry (AMS) performed at the University of Notre Dame's Nuclear Science Laboratory. Samples are prepared by combustion of a small amount of material and subsequent reduction to carbon into an iron powder matrix (graphitization). The graphitization procedure affects the maximum measurement rate, and a poor graphitization can be detrimental to the AMS measurement of the sample. Previous graphitization procedures resulted in a particle current too low or inconsistent to optimize AMS measurements. Thus, there was a desire to design and refine the graphitization system. The finalized process yielded physically darker samples and increased sample currents by two orders of magnitude. Additionally, the first testing of the samples was successful, yet analysis of the dates proved inconclusive. AMS measurements will be performed again to obtain better sampling statistics in the hopes of narrowing the reported date ranges. NSF and JINA-CEE.

Organic matter and containment of uranium and fissiogenic isotopes at the Oklo natural reactors

USGS Publications Warehouse

Nagy, B.; Gauthier-Lafaye, F.; Holliger, P.; Davis, D.W.; Mossman, D.J.; Leventhal, J.S.; Rigali, M.J.; Parnell, J.

1991-01-01

SOME of the Precambrian natural fission reactors at Oklo in Gabon contain abundant organic matter1,2, part of which was liquefied at the time of criticality and subsequently converted to a graphitic solid3,4. The liquid organic matter helps to reduce U(VI) to U(IV) from aqueous solutions, resulting in the precipitation of uraninite5. It is known that in the prevailing reactor environments, precipitated uraninite grains incorporated fission products. We report here observations which show that these uraninite crystals were held immobile within the resolidified, graphitic bitumen. Unlike water-soluble (humic) organic matter, the graphitic bituminous organics at Oklo thus enhanced radionu-clide containment. Uraninite encased in solid graphitic matter in the organic-rich reactor zones lost virtually no fissiogenic lan-thanide isotopes. The first major episode of uranium and lead migration was caused by the intrusion of a swarm of adjacent dolerite dykes about 1,100 Myr after the reactors went critical. Our results from Oklo imply that the use of organic, hydrophobic solids such as graphitic bitumen as a means of immobilizing radionuclides in pretreated nuclear waste warrants further investigation. ?? 1991 Nature Publishing Group.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Davenport, Michael; Petti, D. A.

The United States Department of Energy’s Advanced Reactor Technologies (ART) Program will irradiate up to six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments are being irradiated over an approximate eight year period to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Very High Temperature Gasmore » Reactor (VHTR), as well as other future gas reactors. The experiments each consist of a single capsule that contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens are not be subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There are also samples taken of the sweep gas effluent to measure any oxidation or off-gassing of the specimens that may occur during initial start-up of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. The third experiment, AGC-3, started its irradiation in late November 2012 and completed in the April of 2014. AGC-4 is currently being irradiated in the ATR. This paper will briefly discuss the preliminary irradiation results of the AGC-4 experiment, as well as the design of AGC-5.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pinhero, Patrick; Windes, William

2015-03-10

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

Fluid-deposited graphitic inclusions in quartz: Comparison between KTB (German Continental Deep-Drilling) core samples and artificially reequilibrated natural inclusions

USGS Publications Warehouse

Pasteris, J.D.; Chou, I.-Ming

1998-01-01

We used Raman microsampling spectroscopy (RMS) to determine the degree of crystallinity of minute (2-15 ??m) graphite inclusions in quartz in two sets of samples: experimentally reequilibrated fluid inclusions in a natural quartz grain and biotite-bearing paragneisses from the KTB deep drillhole in SE Germany. Our sequential reequilibration experiments at 725??C on initially pure CO2 inclusions in a quartz wafer and the J. Krautheim (1993) experiments at 900-1100??C on organic compounds heated in gold or platinum capsules suggest that, at a given temperature, (1) fluid-deposited graphite will have a lower crystallinity than metamorphosed organic matter and (2) that the crystallinity of fluid-deposited graphite is affected by the composition of the fluid from which it was deposited. We determined that the precipitation of more-crystalline graphite is favored by lower fH2 (higher fO2), and that the crystallinity of graphite is established by the conditions (including gas fugacities) that pertain as the fluid first reaches graphite saturation. Graphite inclusions within quartz grains in the KTB rocks show a wide range in crystallinity index, reflecting three episodes of carbon entrapment under different metamorphic conditions. Isolated graphite inclusions have the spectral properties of totally ordered, completely crystalline graphite. Such crystallinity suggests that the graphite was incorporated from the surrounding metasedimentary rocks, which underwent metamorphism at upper amphibolite-facies conditions. Much of the fluid-deposited graphite in fluid inclusions, however, shows some spectral disorder. The properties of that graphite resemble those of experimental precipitates at temperatures in excess of 700??C and at elevated pressures, suggesting that the inclusions represent precipitates from C-O-H fluids trapped under conditions near those of peak metamorphism at the KTB site. In contrast, graphite that is intimately associated with chlorite and other (presumably low-temperature) silicates in inclusions is highly disordered and spectrally resembles kerogens. This graphite probably was deposited during later greenschist-facies retrograde metamorphism at about 400-500??C. The degree of crystallinity of fluid-deposited graphite is shown to be a much more complex function of temperature than is the crystallinity of metamorphic graphite. To some extent, experiments can provide temperature-calibration of the crystallinity index. However, the difference in time scales between experimental runs and geologic processes makes it difficult to infer specific temperatures for naturally precipitated graphite. Copyright ?? 1998 Elsevier Science Ltd.

METAL SHEATHED BODIES

DOEpatents

Finniston, H.M.; Wyatt, L.M.; Plail, O.S.

1961-06-27

An aluminum-cased uranium fuel element is patented for use in nuclear reactors. A layer of a substance such as graphite or a metallic film, preferably of relatively low thermal-neutron capture cross section, between the uranium and aluminum prevents their interdiffusion.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zeng, Fan W.; Contescu, Cristian I.; Gallego, Nidia C.

Laser ultrasonic line source methods have been used to study elastic anisotropy in nuclear graphites by measuring shear wave birefringence. Depending on the manufacturing processes used during production, nuclear graphites can exhibit various degrees of material anisotropy related to preferred crystallite orientation and to microcracking. In this paper, laser ultrasonic line source measurements of shear wave birefringence on NBG-25 have been performed to assess elastic anisotropy. Laser line sources allow specific polarizations for shear waves to be transmitted – the corresponding wavespeeds can be used to compute bulk, elastic moduli that serve to quantify anisotropy. These modulus values can bemore » interpreted using physical property models based on orientation distribution coefficients and microcrack-modified, single crystal moduli to represent the combined effects of crystallite orientation and microcracking on material anisotropy. Finally, ultrasonic results are compared to and contrasted with measurements of anisotropy based on the coefficient of thermal expansion to show the relationship of results from these techniques.« less

Nuclear reactor shield including magnesium oxide

DOEpatents

Rouse, Carl A.; Simnad, Massoud T.

1981-01-01

An improvement in nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux, the reactor shielding including means providing structural support, neutron moderator material, neutron absorber material and other components as described below, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron.

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

NASA Astrophysics Data System (ADS)

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

2018-02-01

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

Supernova Dust at Sub-micrometer Scales

NASA Astrophysics Data System (ADS)

Nittler, Larry; Stroud, R. M.

2006-06-01

Meteorites contain nanometer to micrometer stardust grains, which formed in pre-solar generations of stars and exhibit large isotopic anomalies that reflect the nuclear processes that occurred in their individual parent stars [1]. Supernovae of Type II have been identified as the sources of much of the stardust, including grains of SiC, Si3N4, graphite and Mg2SiO4. Although, the isotopic compositions of many elements in these grains point unambiguously to supernova nucleosynthesis processes [2], the data require extensive and heterogeneous mixing of disparate nuclear burning zones. A recent study found that individual 200 nm TiC sub-grains within a 12 micron supernova graphite grain have uniform Ti isotopic composition but a range of O isotopic ratios [3]. New microanalysis techniques allow us to correlate the physical microstructures of supernova grains with isotopic composition, e.g., SiC and Si3N4, providing a sub-micron view of condensation processes in supernova ejecta. Results on two SiC grains indicate that micron-sized SiC grains from supernovae consist of assemblages of smaller crystallites with some evidence of radiation and/or shock processing. This is in strong contrast to SiC grains from AGB stars, which are typically single euhedral crystals [4]. The Si, C and N isotopic compositions of the grains are highly uniform, in contrast to the model of [5], which predicts strong isotopic gradients in supernova-derived SiC grains.This work is supported by NASA.[1] Clayton D. D. and Nittler L. R. (2004) ARAA, 42, 39-78.[2] Nittler L. R., et al. (1996) ApJ, 462, L31-34.[3] Stadermann F. J., et al. (2005) GCA, 69, 177-188.[4] Daulton T. L., et al. (2002) Science, 296, 1852-1855.[5] Deneault E. A.-N., et al. (2003) ApJ, 594, 312-325.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Braun,A.; Huggins, F.; Kelly, K.

We report on the structure of a set of diesel exhaust samples that were obtained from reference diesel fuel and diesel fuel mixed with ferrocene. Characterization was carried out with X-ray absorption spectroscopy (C(1s) NEXAFS) and wide-angle X-ray scattering (WAXS). The reference diesel soot shows a pronounced graphite-like microstructure and molecular structure, with a strong (0 0 2) graphite Bragg reflex and a strong aromatic C{double_bond}C resonance at 285 eV. The mineral matter in the reference soot could be identified as Fe{sub 2}O{sub 3} hematite. The soot specimen from the diesel mixed with ferrocene has an entirely different structure andmore » lacks significantly in graphite-like characteristics. NEXAFS spectra of such soot barely show aromatics but pronounced contributions from aliphatic structures. WAXS patterns show almost no intensity at the Bragg (0 0 2) reflection of graphite, but a strong aliphatic {gamma}-side band. The iron from the ferrocene transforms to Fe{sub 2}O{sub 3} maghemite.« less

The 13C nuclear magnetic resonance in graphite intercalation compounds

NASA Technical Reports Server (NTRS)

Tsang, T.; Resing, H. A.

1985-01-01

The (13)C NMR chemical shifts of graphite intercalation compounds were calculated. For acceptor types, the shifts come mainly from the paramagnetic (Ramsey) intra-atomic terms. They are related to the gross features of the two-dimensional band structures. The calculated anisotropy is about -140 ppm and is independent of the finer details such as charge transfer. For donor types, the carbon 2p pi orbitals are spin-polarized because of mixing with metal conduction electrons, thus there is an additional dipolar contribution which may be correlated with the electronic specific heat. The general agreement with experimental data is satisfactory.

C-13 nuclear magnetic resonance in graphite intercalation compounds

NASA Technical Reports Server (NTRS)

Tsang, T.; Resing, H. A.

1985-01-01

The C-13 NMR chemical shifts of graphite intercalation compounds have been calculated. For acceptor types, the shifts come mainly from the paramagnetic (Ramsey) intra-atomic terms. They are related to the gross features of the two-dimensional band structures. The calculated anisotropy is about - 140 ppm and is independent of the finer details such as charge transfer. For donor types, the carbon 2p pi orbitals are spin-polarized because of mixing with metal-conduction electrons, thus there is an additional dipolar contribution which may be correlated with the electronic specific heat. The general agreement with experimental data is satisfactory.

Apparatus for controlling molten core debris

DOEpatents

Golden, Martin P. [Trafford, PA; Tilbrook, Roger W. [Monroeville, PA; Heylmun, Neal F. [Pittsburgh, PA

1977-07-19

Apparatus for containing, cooling, diluting, dispersing and maintaining subcritical the molten core debris assumed to melt through the bottom of a nuclear reactor pressure vessel in the unlikely event of a core meltdown. The apparatus is basically a sacrificial bed system which includes an inverted conical funnel, a core debris receptacle including a spherical dome, a spherically layered bed of primarily magnesia bricks, a cooling system of zig-zag piping in graphite blocks about and below the bed and a cylindrical liner surrounding the graphite blocks including a steel shell surrounded by firebrick. Tantalum absorber rods are used in the receptacle and bed.

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

DOE PAGES

Lai, Shigang; Shi, Li; Fok, Alex; ...

2017-01-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lai, Shigang; Shi, Li; Fok, Alex

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

Durability of Intercalated Graphite Epoxy Composites in Low Earth Orbit

NASA Technical Reports Server (NTRS)

Gaier, James R.; Davidson, Michelle L.; Shively, Rhonda

1996-01-01

The electrical conductivity of graphite epoxy composites can be substantially increased by intercalating (inserting guest atoms or molecules between the graphene planes) the graphite fibers before composite formation. The resulting high strength, low density, electrically conducting composites have been proposed for EMI shielding in spacecraft. Questions have been raised, however, about their durability in the space environment, especially with respect to outgassing of the intercalates, which are corrosive species such as bromine. To answer those concerns, six samples of bromine intercalated graphite epoxy composites were included in the third Evaluation of Oxygen Interaction with Materials (EOIM-3) experiment flown on the Space Shuttle Discovery (STS-46). Changes in electrical conductivity, optical reflectance, surface texture, and mass loss for SiO2 protected and unprotected samples were measured after being exposed to the LEO environment for 42 hours. SiO2 protected samples showed no degradation, verifying conventional protection strategies are applicable to bromine intercalated composites. The unprotected samples showed that bromine intercalation does not alter the degradation of graphite-epoxy composites. No bromine was detected to have been released by the fibers allaying fears that outgassing could be disruptive to the sensitive electronics the EMI shield is meant to protect.

Reflection of Low Energy Positrons from the Surface of Highly Oriented Pyrolytic Graphite and Single Layer Graphene.

NASA Astrophysics Data System (ADS)

Imam, S. K.; Chirayath, V. A.; Chrysler, M. D.; Fairchild, A. J.; Gladen, R. W.; Koymen, A. R.; Weiss, A. H.; UT Arlington Positron Surface Laboratory Team

A time of flight positron annihilation induced Auger electron spectrometer (TOF-PAES) was utilized to measure the reflection of positrons as a function of incident positron energy (0 to 10 eV) from the surface of highly oriented pyrolytic graphite (HOPG) and from a single layer graphene (SLG) on a Cu foil. A NaI scintillation detector was used to measure the annihilation gamma from the reflected positrons as a function of incident positron kinetic energy. The annihilation of the positrons on HOPG and SLG were simultaneously measured using another NaI detector near the sample. The Auger electrons emitted as a result of the annihilation of positrons from the surface of the sample were also measured concurrently. As the positron kinetic energy was increased, the number of reflected positrons calculated from the intensity under the annihilation gamma peak showed a steady decrease. The positronium formation measured at the sample using the gamma spectrum showed a peak at 6 eV. The intensity of the carbon KVV Auger peak showed a dip at the same energy. The correlation of the three signals, intensity of reflected positrons, positrons annihilating at the sample and the Auger intensity are discussed for both samples. This work was supported by NSF Grant No. DMR 1508719 and DMR 1338130.

Fluence correction factors for graphite calorimetry in a low-energy clinical proton beam: I. Analytical and Monte Carlo simulations.

PubMed

Palmans, H; Al-Sulaiti, L; Andreo, P; Shipley, D; Lühr, A; Bassler, N; Martinkovič, J; Dobrovodský, J; Rossomme, S; Thomas, R A S; Kacperek, A

2013-05-21

The conversion of absorbed dose-to-graphite in a graphite phantom to absorbed dose-to-water in a water phantom is performed by water to graphite stopping power ratios. If, however, the charged particle fluence is not equal at equivalent depths in graphite and water, a fluence correction factor, kfl, is required as well. This is particularly relevant to the derivation of absorbed dose-to-water, the quantity of interest in radiotherapy, from a measurement of absorbed dose-to-graphite obtained with a graphite calorimeter. In this work, fluence correction factors for the conversion from dose-to-graphite in a graphite phantom to dose-to-water in a water phantom for 60 MeV mono-energetic protons were calculated using an analytical model and five different Monte Carlo codes (Geant4, FLUKA, MCNPX, SHIELD-HIT and McPTRAN.MEDIA). In general the fluence correction factors are found to be close to unity and the analytical and Monte Carlo codes give consistent values when considering the differences in secondary particle transport. When considering only protons the fluence correction factors are unity at the surface and increase with depth by 0.5% to 1.5% depending on the code. When the fluence of all charged particles is considered, the fluence correction factor is about 0.5% lower than unity at shallow depths predominantly due to the contributions from alpha particles and increases to values above unity near the Bragg peak. Fluence correction factors directly derived from the fluence distributions differential in energy at equivalent depths in water and graphite can be described by kfl = 0.9964 + 0.0024·zw-eq with a relative standard uncertainty of 0.2%. Fluence correction factors derived from a ratio of calculated doses at equivalent depths in water and graphite can be described by kfl = 0.9947 + 0.0024·zw-eq with a relative standard uncertainty of 0.3%. These results are of direct relevance to graphite calorimetry in low-energy protons but given that the fluence correction factor is almost solely influenced by non-elastic nuclear interactions the results are also relevant for plastic phantoms that consist of carbon, oxygen and hydrogen atoms as well as for soft tissues.

Development of composite facets for the surface of a space-based solar dynamic concentrator

NASA Technical Reports Server (NTRS)

Ayers, Schuyler R.; Morel, Donald E.; Sanborn, James A.

1986-01-01

An account is given of the composite fabrication techniques envisioned for the production of mirror-quality substrates furnishing the specular reflectance required for the NASA Space Station's solar dynamic concentrator energy system. The candidate materials were graphite fiber-reinforced glass, aluminum, and polymer matrices whose surfaces would be coated with thin metal layers and with atomic oxygen degradation-inhibiting protective coatings to obtain the desired mirror surface. Graphite-epoxy mirror substrate samples have been found to perform satisfactorily for the required concentrator lifetime.

A new ring-shaped graphite monitor ionization chamber

NASA Astrophysics Data System (ADS)

Yoshizumi, M. T.; Caldas, L. V. E.

2010-07-01

A ring-shaped monitor ionization chamber was developed at the Instituto de Pesquisas Energéticas e Nucleares. This ionization chamber presents an entrance window of aluminized polyester foil. The guard ring and collecting electrode are made of graphite coated Lucite plates. The main difference between this new ionization chamber and commercial monitor chambers is its ring-shaped design. The new monitor chamber has a central hole, allowing the passage of the direct radiation beam without attenuation; only the penumbra radiation is measured by the sensitive volume. This kind of ionization chamber design has already been tested, but using aluminium electrodes. By changing the electrode material from aluminium to a graphite coating, an improvement in the chamber response stability was expected. The pre-operational tests, as saturation curve, recombination loss and polarity effect showed satisfactory results. The repeatability and the long-term stability tests were also evaluated, showing good agreement with international recommendations.

Ultra high vacuum adhesion testing of NERVA engine materials

NASA Technical Reports Server (NTRS)

1970-01-01

The primary objective of this research program was to determine the effects of surface cleaning and deliberate gaseous contamination on the adhesion behavior of selected candidate materials for use in the NERVA nuclear rocket engine program. Using a torsion balance technique, the relationship between the normal compressive load applied to crossed rod samples and the resultant contact resistance was used to ascertain the extent of adhesion under each set of experimental conditions. In addition to an evaluation of the static adhesion behavior of selected materials combinations, the experimental apparatus was modified to permit a similar investigation relating to the effects of specific tangential displacements of the sample wires, i.e., their sliding friction behavior. During the course of this subcontract, the materials combinations 440 C vs. 440 C. pyrographite vs ZTA graphite, Nbc (graphite) vs. Nbc (graphite), and Electrolize Inconel 718 vs. Au electroplated 302 S/S were evaluated.

Effect of specimen size and grain orientation on the mechanical and physical properties of NBG-18 nuclear graphite

DOE PAGES

Vasudevamurthy, G.; Byun, T. S.; Pappano, Pete; ...

2015-03-13

Here we present a comparison of the measured baseline mechanical and physical properties of with grain (WG) and against grain (AG) non-ASTM size NBG-18 graphite. The objectives of the experiments were twofold: (1) assess the variation in properties with grain orientation; (2) establish a correlation between specimen tensile strength and size. The tensile strength of the smallest sized (4 mm diameter) specimens were about 5% higher than the standard specimens (12 mm diameter) but still within one standard deviation of the ASTM specimen size indicating no significant dependence of strength on specimen size. The thermal expansion coefficient and elastic constantsmore » did not show significant dependence on specimen size. Lastly, experimental data indicated that the variation of thermal expansion coefficient and elastic constants were still within 5% between the different grain orientations, confirming the isotropic nature of NBG-18 graphite in physical properties.« less

Conversion from dose-to-graphite to dose-to-water in an 80 MeV/A carbon ion beam.

PubMed

Rossomme, S; Palmans, H; Shipley, D; Thomas, R; Lee, N; Romano, F; Cirrone, P; Cuttone, G; Bertrand, D; Vynckier, S

2013-08-21

Based on experiments and numerical simulations, a study is carried out pertaining to the conversion of dose-to-graphite to dose-to-water in a carbon ion beam. This conversion is needed to establish graphite calorimeters as primary standards of absorbed dose in these beams. It is governed by the water-to-graphite mass collision stopping power ratio and fluence correction factors, which depend on the particle fluence distributions in each of the two media. The paper focuses on the experimental and numerical determination of this fluence correction factor for an 80 MeV/A carbon ion beam. Measurements have been performed in the nuclear physics laboratory INFN-LNS in Catania (Sicily, Italy). The numerical simulations have been made with a Geant4 Monte Carlo code through the GATE simulation platform. The experimental data are in good agreement with the simulated results for the fluence correction factors and are found to be close to unity. The experimental values increase with depth reaching 1.010 before the Bragg peak region. They have been determined with an uncertainty of 0.25%. Different numerical results are obtained depending on the level of approximation made in calculating the fluence correction factors. When considering carbon ions only, the difference between measured and calculated values is maximal just before the Bragg peak, but its value is less than 1.005. The numerical value is close to unity at the surface and increases to 1.005 near the Bragg peak. When the fluence of all charged particles is considered, the fluence correction factors are lower than unity at the surface and increase with depth up to 1.025 before the Bragg peak. Besides carbon ions, secondary particles created due to nuclear interactions have to be included in the analysis: boron ions ((10)B and (11)B), beryllium ions ((7)Be), alpha particles and protons. At the conclusion of this work, we have the conversion of dose-to-graphite to dose-to-water to apply to the response of a graphite calorimeter in an 80 MeV/A carbon ion beam. This conversion consists of the product of two contributions: the water-to-graphite electronic mass collision stopping power ratio, which is equal to 1.115, and the fluence correction factor which varies linearly with depth, as k(fl, all) = 0.9995 + 0.0048(zw-eq). The latter has been determined on the basis of experiments and numerical simulations.

Teleoperated systems for nuclear reactors: Inspection and maintenance

NASA Technical Reports Server (NTRS)

Dorokhov, V. P.; Dorokhov, D. V.; Eperin, A. P.

1994-01-01

The present paper describes author's work in the field of teleoperated equipment for inspection and maintenance of the RBML technological channels and graphite laying, emergency operations. New technological and design solutions of teleoperated robotic systems developed for Leningradsky Power Plant are discussed.

Multi-Physics Simulation of TREAT Kinetics using MAMMOTH

DOE Office of Scientific and Technical Information (OSTI.GOV)

DeHart, Mark; Gleicher, Frederick; Ortensi, Javier

With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific fuels transient tests range from simple temperature transients to full fuel melt accidents. The current TREAT core is driven by highly enriched uranium (HEU) dispersed in amore » graphite matrix (1:10000 U-235/C atom ratio). At the center of the core, fuel is removed allowing for the insertion of an experimental test vehicle. TREAT’s design provides experimental flexibility and inherent safety during neutron pulsing. This safety stems from the graphite in the driver fuel having a strong negative temperature coefficient of reactivity resulting from a thermal Maxwellian shift with increased leakage, as well as graphite acting as a temperature sink. Air cooling is available, but is generally used post-transient for heat removal. DOE and INL have expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility, with an emphasis on effective and safe operation while minimizing experimental time and cost. At INL, the Multi-physics Object Oriented Simulation Environment (MOOSE) has been selected as the model development framework for this work. This paper describes the results of preliminary simulations of a TREAT fuel element under transient conditions using the MOOSE-based MAMMOTH reactor physics tool.« less

DR Reactor VSR channel damage

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kempf, F.J.; Rawlins, J.K.

1961-10-30

On July 11, 1961 the Ball 3X System at DR Reactor was inadventently tripped. All vertical safety rods dropped and all channels were filled with balls. This report has the twofold purpose of documenting borescope observations of ten vertical rod channels at DR Reactor and recording the estimated extent of graphite damage resulting from the above incident. Channel damage data are presented on appended drawings. With suitable notations, the tracings of these drawings may be revised to reflect any future graphite damage. All vertical rod channels at DR Reactor were visually examined with a closed circuit television system during ballmore » removal efforts. Typical photographs of trapped balls and ledges, as viewed on the television monitor, are shown. Photographs of typical graphite damage, obtained through the borescope are also included in this report. 3 refs., 8 figs., 1 tab.« less

Surface Anchoring of Nematic Phase on Carbon Nanotubes: Nanostructure of Ultra-High Temperature Materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ogale, Amod A

2012-04-27

Nuclear energy is a dependable and economical source of electricity. Because fuel supply sources are available domestically, nuclear energy can be a strong domestic industry that can reduce dependence on foreign energy sources. Commercial nuclear power plants have extensive security measures to protect the facility from intruders [1]. However, additional research efforts are needed to increase the inherent process safety of nuclear energy plants to protect the public in the event of a reactor malfunction. The next generation nuclear plant (NGNP) is envisioned to utilize a very high temperature reactor (VHTR) design with an operating temperature of 650-1000°C [2]. Onemore » of the most important safety design requirements for this reactor is that it must be inherently safe, i.e., the reactor must shut down safely in the event that the coolant flow is interrupted [2]. This next-generation Gen IV reactor must operate in an inherently safe mode where the off-normal temperatures may reach 1500°C due to coolant-flow interruption. Metallic alloys used currently in reactor internals will melt at such temperatures. Structural materials that will not melt at such ultra-high temperatures are carbon/graphtic fibers and carbon-matrix composites. Graphite does not have a measurable melting point; it is known to sublime starting about 3300°C. However, neutron radiation-damage effects on carbon fibers are poorly understood. Therefore, the goal of this project is to obtain a fundamental understanding of the role of nanotexture on the properties of resulting carbon fibers and their neutron-damage characteristics. Although polygranular graphite has been used in nuclear environment for almost fifty years, it is not suitable for structural applications because it do not possess adequate strength, stiffness, or toughness that is required of structural components such as reaction control-rods, upper plenum shroud, and lower core-support plate [2,3]. For structural purposes, composites consisting of strong carbon fibers embedded in a carbon matrix are needed. Such carbon/carbon (C/C) composites have been used in aerospace industry to produce missile nose cones, space shuttle leading edge, and aircraft brake-pads. However, radiation-tolerance of such materials is not adequately known because only limited radiation studies have been performed on C/C composites, which suggest that pitch-based carbon fibers have better dimensional stability than that of polyacrylonitrile (PAN) based fibers [4]. The thermodynamically-stable state of graphitic crystalline packing of carbon atoms derived from mesophase pitch leads to a greater stability during neutron irradiation [5]. The specific objectives of this project were: (i) to generating novel carbonaceous nanostructures, (ii) measure extent of graphitic crystallinity and the extent of anisotropy, and (iii) collaborate with the Carbon Materials group at Oak Ridge National Lab to have neutron irradiation studies and post-irradiation examinations conducted on the carbon fibers produced in this research project.« less

Apparatus for controlling molten core debris. [LMFBR

DOEpatents

Golden, M.P.; Tilbrook, R.W.; Heylmun, N.F.

1977-07-19

Disclosed is an apparatus for containing, cooling, diluting, dispersing and maintaining subcritical the molten core debris assumed to melt through the bottom of a nuclear reactor pressure vessel in the unlikely event of a core meltdown. The apparatus is basically a sacrificial bed system which includes an inverted conical funnel, a core debris receptacle including a spherical dome, a spherically layered bed of primarily magnesia bricks, a cooling system of zig-zag piping in graphite blocks about and below the bed and a cylindrical liner surrounding the graphite blocks including a steel shell surrounded by firebrick. Tantalum absorber rods are used in the receptacle and bed. 9 claims, 22 figures.

Benchmarking of HEU Mental Annuli Critical Assemblies with Internally Reflected Graphite Cylinder

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xiaobo, Liu; Bess, John D.; Marshall, Margaret A.

Three experimental configurations of critical assemblies, performed in 1963 at the Oak Ridge Critical Experiment Facility, which are assembled using three different diameter HEU annuli (15-9 inches, 15-7 inches and 13-7 inches) metal annuli with internally reflected graphite cylinder are evaluated and benchmarked. The experimental uncertainties which are 0.00055, 0.00055 and 0.00055 respectively, and biases to the detailed benchmark models which are -0.00179, -0.00189 and -0.00114 respectively, were determined, and the experimental benchmark keff results were obtained for both detailed and simplified model. The calculation results for both detailed and simplified models using MCNP6-1.0 and ENDF VII.1 agree well tomore » the benchmark experimental results with a difference of less than 0.2%. These are acceptable benchmark experiments for inclusion in the ICSBEP Handbook.« less

Neutron transmission measurements of poly and pyrolytic graphite crystals

NASA Astrophysics Data System (ADS)

Adib, M.; Abbas, Y.; Abdel-Kawy, A.; Ashry, A.; Kilany, M.; Kenawy, M. A.

The total neutron cross-section measurements of polycrystalline graphite have been carried out in a neutron wavelength from 0.04 to 0.78 nm. This work also presents the neutron transmission measurements of pyrolytic graphite (PG) crystal in a neutron wavelength band from 0.03 to 0.50 nm, at different orientations of the PG crystal with regard to the beam direction. The measurements were performed using three time-of-flight (TOF) spectrometers installed in front of three of the ET-RR-1 reactor horizontal channels. The average value of the coherent scattering amplitude for polycrystalline graphite was calculated and found to be bcoh = (6.61 ± 0.07) fm. The behaviour of neutron transmission through the PG crystal, while oriented at different angles with regard to the beam direction, shows dips at neutron wavelengths corresponding to the reflections from (hkl) planes of hexagonal graphite structure. The positions of the observed dips are found to be in good agreement with the calculated ones. It was also found that a 40 mm thick PG crystal is quite enough to reduce the second-order contamination of the neutron beam from 2.81 to 0.04, assuming that the incident neutrons have a Maxwell distribution with neutron gas temperature 330 K.

What makes lithium substituted polyacrylic acid a better binder than polyacrylic acid for silicon-graphite composite anodes?

DOE PAGES

Hays, Kevin A.; Ruther, Rose E.; Kukay, Alexander J.; ...

2018-03-04

Lithium substituted polyacrylic acid (LiPAA) has previously been demonstrated as a superior binder over polyacrylic acid (PAA) for Si anodes, but from where does this enhanced performance arise? In this paper, full cells are assembled with PAA and LiPAA based Si-graphite composite anodes that dried at temperatures from 100 °C to 200 °C. The performance of full cells containing PAA based Si-graphite anodes largely depend on the secondary drying temperature, as decomposition of the binder is correlated to increased electrode moisture and a rise in cell impedance. Full cells containing LiPAA based Si-graphite composite electrodes display better Coulombic efficiency thanmore » those with PAA, because of the electrochemical reduction of the PAA binder. This is identified by attenuated total reflectance Fourier transform infrared spectrometry and observed gassing during the electrochemical reaction. Finally, Coulombic losses from the PAA and Si SEI, along with depletion of the Si capacity in the anode results in progressive underutilization of the cathode and full cell capacity loss.« less

What makes lithium substituted polyacrylic acid a better binder than polyacrylic acid for silicon-graphite composite anodes?

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hays, Kevin A.; Ruther, Rose E.; Kukay, Alexander J.

Lithium substituted polyacrylic acid (LiPAA) has previously been demonstrated as a superior binder over polyacrylic acid (PAA) for Si anodes, but from where does this enhanced performance arise? In this paper, full cells are assembled with PAA and LiPAA based Si-graphite composite anodes that dried at temperatures from 100 °C to 200 °C. The performance of full cells containing PAA based Si-graphite anodes largely depend on the secondary drying temperature, as decomposition of the binder is correlated to increased electrode moisture and a rise in cell impedance. Full cells containing LiPAA based Si-graphite composite electrodes display better Coulombic efficiency thanmore » those with PAA, because of the electrochemical reduction of the PAA binder. This is identified by attenuated total reflectance Fourier transform infrared spectrometry and observed gassing during the electrochemical reaction. Finally, Coulombic losses from the PAA and Si SEI, along with depletion of the Si capacity in the anode results in progressive underutilization of the cathode and full cell capacity loss.« less

What makes lithium substituted polyacrylic acid a better binder than polyacrylic acid for silicon-graphite composite anodes?

NASA Astrophysics Data System (ADS)

Hays, Kevin A.; Ruther, Rose E.; Kukay, Alexander J.; Cao, Pengfei; Saito, Tomonori; Wood, David L.; Li, Jianlin

2018-04-01

Lithium substituted polyacrylic acid (LiPAA) has previously been demonstrated as a superior binder over polyacrylic acid (PAA) for Si anodes, but from where does this enhanced performance arise? In this study, full cells are assembled with PAA and LiPAA based Si-graphite composite anodes that dried at temperatures from 100 °C to 200 °C. The performance of full cells containing PAA based Si-graphite anodes largely depend on the secondary drying temperature, as decomposition of the binder is correlated to increased electrode moisture and a rise in cell impedance. Full cells containing LiPAA based Si-graphite composite electrodes display better Coulombic efficiency than those with PAA, because of the electrochemical reduction of the PAA binder. This is identified by attenuated total reflectance Fourier transform infrared spectrometry and observed gassing during the electrochemical reaction. Coulombic losses from the PAA and Si SEI, along with depletion of the Si capacity in the anode results in progressive underutilization of the cathode and full cell capacity loss.

Identification and Quantification of Carbon Phases in Conversion Fuel for the Transient Reactor Test Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steele, Robert; Mata, Angelica; Dunzik-Gougar, Mary Lou

2016-06-01

As part of an overall effort to convert US research reactors to low-enriched uranium (LEU) fuel use, a LEU conversion fuel is being designed for the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory. TREAT fuel compacts are comprised of UO2 fuel particles in a graphitic matrix material. In order to refine heat transfer modeling, as well as determine other physical and nuclear characteristics of the fuel, the amount and type of graphite and non-graphite phases within the fuel matrix must be known. In this study, we performed a series of complementary analyses, designed to allow detailed characterizationmore » of the graphite and phenolic resin based fuel matrix. Methods included Scanning Electron and Transmission Electron Microscopies, Raman spectroscopy, X-ray Diffraction, and Dual-Beam Focused Ion Beam Tomography. Our results indicate that no single characterization technique will yield all of the desired information; however, through the use of statistical and empirical data analysis, such as curve fitting, partial least squares regression, volume extrapolation and spectra peak ratios, a degree of certainty for the quantity of each phase can be obtained.« less

Biomimetic graphene sensors: functionalizing graphene with peptides

NASA Astrophysics Data System (ADS)

Ishigami, Masa; Nyon Kim, Sang; Naik, Rajesh; Tatulian, Suren A.; Katoch, Jyoti

2012-02-01

Non-covalent biomimetic functionalization of graphene using peptides is one of more promising methods for producing novel sensors with high sensitivity and selectivity. Here we combine atomic force microscopy, Raman spectroscopy, and attenuated total reflection Fourier transform infrared spectroscopy to investigate peptide binding to graphene and graphite. We choose to study a dodecamer peptide identified with phage display to possess affinities for graphite and we find that the peptide forms a complex mesh-like structure upon adsorption on graphene. Moreover, optical spectroscopy reveals that the peptide binds non-covalently to graphene and possesses an optical signature of an ?-helical conformation on graphene.

Ion sputter textured graphite electrode plates

NASA Technical Reports Server (NTRS)

Curren, A. N.; Forman, R.; Sovey, J. S.; Wintucky, E. G. (Inventor)

1983-01-01

A specially textured surface of pyrolytic graphite exhibits extremely low yields of secondary electrons and reduced numbers of reflected primary electrons after impingement of high energy primary electrons. Electrode plates of this material are used in multistage depressed collectors. An ion flux having an energy between 500 iV and 1000 iV and a current density between 1.0 mA/sq cm and 6.0 mA/sq cm produces surface roughening or texturing which is in the form of needles or spires. Such textured surfaces are especially useful as anode collector plates in high tube devices.

Synthesis and characterization of CdS-based ternary composite for enhanced visible light-driven photocatalysis

NASA Astrophysics Data System (ADS)

Singh, Arvind; Sinha, A. S. K.

2018-09-01

Active ternary graphite and alumina-supported cadmium sulphide (CdS) composite was synthesized by impregnation method followed by high-temperature solid-gas reaction and characterized by X-ray diffraction (XRD), photoluminescence spectroscopy (PL), diffuse reflectance spectroscopy (DRS), scanning electron microscopy (SEM), transmission electron microscopy (TEM), energy-dispersive X-ray spectroscopy (EDX) and X-ray photoelectron spectroscopy (XPS) techniques. The ternary CdS-graphite-alumina composite exhibited superior catalytic activity compared with the binary CdS-alumina composite due to its better visible-light absorption and higher charge separation. The ternary composite has a bed-type structure. It permits a greater interaction at the interface due to intimate contact between CdS and graphite in the ternary composite. This composite has a highly efficient visible light-driven photocatalytic activity for sustainable hydrogen production. It is also capable of degrading organic dyes in wastewater.

Proceedings of the 13th biennial conference on carbon. Extended abstracts and program

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1977-01-01

Properties of carbon are covered including: mechanical and frictional properties; chemical reactivity and surfaces; aerospace applications; carbonization and graphitization; industrial applications; electrical and thermal properties; biomaterials applications; fibers and composites; nuclear applications; activated carbon and adsorption; advances in carbon characterization; and micromechanics and modeling. (GHT)

Proton induced activity in graphite - comparison between measurement and simulation

NASA Astrophysics Data System (ADS)

Kiselev, Daniela; Bergmann, Ryan; Schumann, Dorothea; Talanov, Vadim; Wohlmuther, Michael

2018-06-01

The Paul Scherrer Institut (PSI) operates the Meson production target stations E and M with 590 MeV protons at currents of up to 2.4 mA. Both targets consist of polycrystalline graphite and rotate with 1 Hz due to the high power deposition (40 kW at 2 mA) in Target E. The graphite wheel is regularly exchanged and disposed as radioactive waste after a maximum of 3 to 4 years in operation, which corresponds to about 30 to 40 Ah of proton fluence. For disposal, the nuclide inventory of the long-lived isotopes (T1/2 > 60 d) has to be calculated and reported to the authorities. Measurements of gamma emitters, as well as 3H, 10Be and 14C, were carried out using different techniques. The measured specific activities are compared to Monte Carlo particle transport simulations performed with MCNPX2.7.0 using the BERTINI-DRESNER-RAL (default model in MCNPX2.7.0) and INCL4.6/ABLA07 as nuclear reaction models.

Physical aging in graphite/epoxy composites

NASA Technical Reports Server (NTRS)

Kong, E. S. W.

1983-01-01

Sub-Tg annealing has been found to affect the properties of graphite/epoxy composites. The network epoxy studied was based on the chemistry of tetraglycidyl 4,4'-diamino-diphenyl methane (TGDDM) crosslinked by 4,4'-diamino-diphenyl sulfone (DDS). Differential scanning calorimetry, thermal mechanical analysis, and solid-state cross-polarized magic-angle-spinning nuclear magnetic resonance spectroscopy have been utilized in order to characterize this process of recovery towards thermodynamic equilibrium. The volume and enthalpy recovery as well as the 'thermoreversibility' aspects of the physical aging are discussed. This nonequilibrium and time-dependent behavior of network epoxies are considered in view of the increasingly wide applications of TGDDM-DDS epoxies as matrix materials of structural composites in the aerospace industry.

AIR COOLED NEUTRONIC REACTOR

DOEpatents

Fermi, E.; Szilard, L.

1958-05-27

A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

Different techniques for characterizing single-walled carbon nanotube purity

NASA Astrophysics Data System (ADS)

Yuca, Neslihan; Camtakan, Zeyneb; Karatepe, Nilgün

2013-09-01

Transition-metal catalysts, fullerenes, graphitic carbon, amorphous carbon, and graphite flakes are the main impurities in carbon nanotubes. In this study, we demonstrate an easy and optimum method of cleaning SWCNTs and evaluating their purity. The purification method, which employed oxidative heat treatment followed by 6M HNO3, H2SO4, HNO3:H2SO4 and HCl acid reflux for 6h at 120°C and microwave digestion with 1.5M HNO3 for 0.5h at 210°C which was straightforward, inexpensive, and fairly effective. The purified materials were characterized by thermogravimetric analysis and nuclear techniques such as INAA, XRF and XRD.

Pre-test CFD Calculations for a Bypass Flow Standard Problem

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rich Johnson

The bypass flow in a prismatic high temperature gas-cooled reactor (HTGR) is the flow that occurs between adjacent graphite blocks. Gaps exist between blocks due to variances in their manufacture and installation and because of the expansion and shrinkage of the blocks from heating and irradiation. Although the temperature of fuel compacts and graphite is sensitive to the presence of bypass flow, there is great uncertainty in the level and effects of the bypass flow. The Next Generation Nuclear Plant (NGNP) program at the Idaho National Laboratory has undertaken to produce experimental data of isothermal bypass flow between three adjacentmore » graphite blocks. These data are intended to provide validation for computational fluid dynamic (CFD) analyses of the bypass flow. Such validation data sets are called Standard Problems in the nuclear safety analysis field. Details of the experimental apparatus as well as several pre-test calculations of the bypass flow are provided. Pre-test calculations are useful in examining the nature of the flow and to see if there are any problems associated with the flow and its measurement. The apparatus is designed to be able to provide three different gap widths in the vertical direction (the direction of the normal coolant flow) and two gap widths in the horizontal direction. It is expected that the vertical bypass flow will range from laminar to transitional to turbulent flow for the different gap widths that will be available.« less

Graphite Waste Tank Cleanup and Decontamination under the Marcoule UP1 D and D Program - 13166

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thomasset, Philippe; Chabeuf, Jean-Michel; Thiebaut, Valerie

2013-07-01

The UP1 plant in Marcoule reprocessed nearly 20,000 tons of used natural uranium gas cooled reactor fuel coming from the first generation of civil nuclear reactors in France. During more than 40 years, the decladding operations produced thousands of tons of processed waste, mainly magnesium and graphite fragments. In the absence of a French repository for the graphite waste, the graphite sludge content of the storage pits had to be retrieved and transferred into a newer and safer pit. After an extensive R and D program, the equipment and process necessary for retrieval operations were designed, built and tested. Themore » innovative process is mainly based on the use of two pumps (one to capture and the other one to transfer the sludge) working one after the other and a robotic arm mounted on a telescopic mast. A dedicated process was also set up for the removal of the biggest fragments. The retrieval of the most irradiating fragments was a challenge. Today, the first pit is totally empty and its stainless steel walls have been decontaminated using gels. In the second pit, the sludge retrieval and transfer operations have been almost completed. Most of the non-pumpable graphite fragments has been removed and transferred to a new storage pit. After more than 6 years of operations in sludge retrieval, a lot of experience was acquired from which important 'lessons learned' could be shared. (authors)« less

Reliability Assessment of Graphite Specimens under Multiaxial Stresses

NASA Technical Reports Server (NTRS)

Sookdeo, Steven; Nemeth, Noel N.; Bratton, Robert L.

2008-01-01

An investigation was conducted to predict the failure strength response of IG-100 nuclear grade graphite exposed to multiaxial stresses. As part of this effort, a review of failure criteria accounting for the stochastic strength response is provided. The experimental work was performed in the early 1990s at the Oak Ridge National Laboratory (ORNL) on hollow graphite tubes under the action of axial tensile loading and internal pressurization. As part of the investigation, finite-element analysis (FEA) was performed and compared with results of FEA from the original ORNL report. The new analysis generally compared well with the original analysis, although some discrepancies in the location of peak stresses was noted. The Ceramics Analysis and Reliability Evaluation of Structures Life prediction code (CARES/Life) was used with the FEA results to predict the quadrants I (tensile-tensile) and quadrant IV (compression-tension) strength response of the graphite tubes for the principle of independent action (PIA), the Weibull normal stress averaging (NSA), and the Batdorf multiaxial failure theories. The CARES/Life reliability analysis showed that all three failure theories gave similar results in quadrant I but that in quadrant IV, the PIA and Weibull normal stress-averaging theories were not conservative, whereas the Batdorf theory was able to correlate well with experimental results. The conclusion of the study was that the Batdorf theory should generally be used to predict the reliability response of graphite and brittle materials in multiaxial loading situations.

O the Determination of the Complex Refractive Index of Powdered Materials in the 9 TO 11 Micrometer Spectral Region Utilizing AN Attenuated Total Reflectance Technique.

NASA Astrophysics Data System (ADS)

Gillespie, James Bryce

1982-03-01

A specific method of determining the complex refractive index of powdered materials using attenuated total reflectance (ATR) spectroscopy was investigated. A very precise laser/goniometric ATR system was assembled and applied to powdered samples of carbon blacks, graphite, kaolin clay, quartz, calcite, and sodalime glass beads. The reflectivity data fell into two categories: (1) data representative of a medium having a unique effective refractive index and (2) data representative of a scattering medium having no unique refractive index. Data of the first kind were obtained from all the carbon black, graphite, and kaolin clay samples. The Fahrenfort-Visser solution of the Fresnel equations was applied to the goniometric reflectivity data for these samples to obtain the complex refractive index of these effective media. The complex refractive index obtained in this manner is not that of the bulk material but is instead a value which may be related to the bulk material value through some refractive index mixing rule. A systematic experiment using carbon black of particle size 0.0106 mm diameter was conducted to determine the applicability of several mixture rules for the volume packing fraction range of .2 to .6 which is most often encountered. The Bruggemann effective medium theory produced credible results while the Lorentz-Lorenz rule and the empirical Biot-Arago rule were invalid in this volume packing region. The Bruggemann rule was applied to lampblack, Mogul-L carbon black, graphite, and kaolin clay to obtain the complex refractive indices of these materials from the ATR spectroscopy data. Goniometric reflectivity data representative of an inhomogeneous scattering medium were obtained from all the powdered quartz, powdered calcite, and sodalime glass beads samples. These samples all contained particles with diameters nearly as large as the wavelength. These data demonstrate that the ATR technique, coupled with an effective medium analysis, may be used to obtain optical constants of powdered materials only when the particles are small compared to the wavelength.

Influence of helium atoms on the shear behavior of the fiber/matrix interphase of SiC/SiC composite

NASA Astrophysics Data System (ADS)

Jin, Enze; Du, Shiyu; Li, Mian; Liu, Chen; He, Shihong; He, Jian; He, Heming

2016-10-01

Silicon carbide has many attractive properties and the SiC/SiC composite has been considered as a promising candidate for nuclear structural materials. Up to now, a computational investigation on the properties of SiC/SiC composite varying in the presence of nuclear fission products is still missing. In this work, the influence of He atoms on the shear behavior of the SiC/SiC interphase is investigated via Molecular Dynamics simulation following our recent paper. Calculations are carried out on three dimensional models of graphite-like PyC/SiC interphase and amorphous PyC/SiC interphase with He atoms in different regions (the SiC region, the interface region and the PyC region). In the graphite-like PyC/SiC interphase, He atoms in the SiC region have little influence on the shear strength of the material, while both the shear strength and friction strength may be enhanced when they are in the PyC region. Low concentration of He atoms in the interface region of the graphite-like PyC/SiC interphase increases the shear strength, while there is a reduction of shear strength when the He concentration is high due to the switch of sliding plane. In the amorphous PyC/SiC interphase, He atoms can cause the reduction of the shear strength regardless of the regions that He atoms are located. The presence of He atoms may significantly alter the structure of SiC/SiC in the interface region. The influence of He atoms in the interface region is the most significant, leading to evident shear strength reduction of the amorphous PyC/SiC interphase with increasing He concentration. The behaviors of the interphases at different temperatures are studied as well. The dependence of the shear strengths of the two types of interphases on temperatures is studied as well. For the graphite-like PyC/SiC interphase, it is found strongly related to the regions He atoms are located. Combining these results with our previous study on pure SiC/SiC system, we expect this work may provide new insight into the mechanism of interphase evolution when SiC/SiC is applied as nuclear materials.

Poco Graphite Inc. SuperSiC 0.25m Mirror Cryogenic Test Result

NASA Technical Reports Server (NTRS)

Eng, Ron; Stahl, Phil; Hogue, Bill; Hadaway, James

2004-01-01

SuperSiC, a low areal density material, developed by POCO Graphite, have been used as mirror substrate for high energy lasers, laser radar systems, surveillance, telescopes, scan mirrors and satellites. SuperSiC has excellent thermal properties and cryogenic stability. It exhibits exceptional polishability for reflective optics with high strength, stiffness, and excellent thermal conductivity. A lightweighted 0.2-diameter polished SuperSic mirror was tested at cryogenic temperature at NASMSFC. Optical test results showed 6nm cry0 deformation from ambient to 30 degrees Kelvin and little to no change in its surface figure due to cry0 cycling.

Ultrasonic nondestructive evaluation of graphite epoxy composite laminates

NASA Technical Reports Server (NTRS)

Miller, James G.

1990-01-01

Quantitative ultrasonic techniques are summarized with applications to the measurement of frequency-dependent attenuation and backscatter and to the NDE of composite laminates. Results are listed for the ultrasonic NDE of graphite-epoxy composite laminates including impact and fatigue damage as well as porosity. The methods reviewed include transmission measurements of attenuation, reconstructive tomography based on attenuation, estimating attenuation from backscattered ultrasound, and backscatter approaches. Phase-sensitive and -insensitive detection techniques are mentioned such as phase cancellation at piezoelectric receiving transducers and acoustoelectric effects. The techniques permit the NDE of the parameters listed in inhomogeneous media and provide both images from the transmission mode and in the reflection mode.

Structural and functional polymer-matrix composites for electromagnetic applications

NASA Astrophysics Data System (ADS)

Wu, Junhua

This dissertation addresses the science and technology of functional and structural polymer-matrix composite materials for electromagnetic applications, which include electromagnetic interference (EMI) shielding and low observability (Stealth). The structural composites are continuous carbon fiber epoxy-matrix composites, which are widely used for airframes. The functional composites are composites with discontinuous fillers and in both bulk and coating forms. Through composite structure variation, attractive electromagnetic properties have been achieved. With no degradation of the tensile strength or modulus, the shielding effectiveness of the structural composites has been improved by enhancing multiple reflections through light activation of the carbon fiber. The multiple reflections loss of the electromagnetic wave increases from 1.1 to 10.2 dB at 1.0 GHz due to the activation. Such a large effect of multiple reflections has not been previously reported in any material. The observability of these composites has been lowered by decreasing the electrical conductivity (and hence decreasing the reflection loss) through carbon fiber coating. The incorporation of mumetal, a magnetic alloy particulate filler (28-40 mum size), in a latex paint has been found to be effective for enhancing the shielding only if the electrical resistivity of the resulting composite coating is below 10 O.cm, as rendered by a conductive particulate filler, such as nickel flake (14-20 mum size). This effectiveness (39 dB at 1.0 GHz) is attributed to the absorption of the electromagnetic wave by the mumetal and the nickel flake, with the high conductivity rendered by the presence of the nickel flake resulting in a relatively high reflection loss of 15.5 dB. Without the nickel flake, the mumetal gives only 3 dB of shielding and 1.5 dB of reflection loss at 1.0 GHz. Nickel powder (0.3-0.5 mum size) has been found to be an effective filler for improving the shielding of polyethersulfone (PES) bulk composites. At 13 vol.%, it gives 90 dB of shielding at 1.0 GHz, compared to 46 dB for nickel powder (20-40 mum) and the prior value of 87 dB reported by Shui and Chung for nickel filament (0.4 mum diameter). The minimum filler content for high shielding is 7-13 vol.% for both nickel powders, compared to 3-7 vol.% for nickel filament. Due to the skin effect, a small filler unit size helps the shielding, which is dominated by reflection. Carbon filament (0.1 mum, >100 mum long, >1000 in aspect ratio) is effective for enhancing the shielding effectiveness of a coating made from a water-based colloid that contains graphite particle (0.7-0.8 mum, 22 wt.%) and a starch-type binder. The filament addition increases the shielding from 11 to 20 dB at 1.0 GHz. This increase in shielding is associated with increase in reflectivity and decrease in electrical resistivity. Graphite flake (5 mum) at the same volume proportion is even more effective; its addition increases the shielding from 11 to 28 dB. The combined use of the graphite flake and a low proportion of stainless steel fiber (11 mum diameter, 2 mm long, 180 in aspect ratio) is yet more effective; it increases the shielding from 11 to 34 dB. Alumina particle (5 mum size, 15 vol.%) is effective for increasing the impedance of a coating made from the graphite colloid by 290%, though the shielding effectiveness is reduced from 18 to 11 dB at 1.0 GHz. The high impedance is attractive for MRIcompatible pacemaker leads. The interface between filler and matrix also affects the shielding. Silane treatment of the surface of graphite flake (5 mum) used in the graphite colloid decreases the viscosity (e.g., from 1750 to 1460 CP), but it also decreases the shielding effectiveness (e.g., from 20 to 16 dB at 1 GHz). Ozone treatment gives a similar effect. The decrease of the shielding effectiveness is attributed to the increase in resistivity due to the surface treatment. Measured and calculated values of the reflection loss are comparable, with the measured value lower than the corresponding calculated value, when the resistivity is sufficiently low (e.g., resistivity below 10 O.cm in case of PES-matrix composites) and a strongly magnetic filler such as mumetal is absent. The agreement is better when the skin depth approaches the specimen thickness. The agreement is worse for the latex paint-based composites than the PES-matrix composites, probably due to superior electrical connectivity in the latter.

Production of an impermeable composite of irradiated graphite and glass by hot isostatic pressing as a long term leach resistant waste form

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fachinger, Johannes; Muller, Walter; Marsat, Eric

2013-07-01

Around 250,000 tons of irradiated graphite (i-graphite) exists worldwide and can be considered as a current waste or future waste stream. The largest national i-graphite inventory is located in UK (∼ 100,000 tons) with significant quantities also in Russia and France [5]. Most of the i-graphite remains in the cores of shutdown nuclear reactors including the MAGNOX type in UK and the UNGG in France. Whilst there are still operational power reactors with graphite cores, such as the Russian RBMKs and the AGRs in UK, all of them will reach their end of life during the next two decades. Themore » most common reference waste management option of i-graphite is a wet or dry retrieval of the graphite blocks from the reactor core and the grouting of these blocks in a container without further conditioning. This produces large waste package volumes because the encapsulation capacity of the grout is limited and large cavities in the graphite blocks could reduce the packing densities. Packing densities from 0.5 to 1 tons per cubic meter have been assumed for grouting solutions. Furthermore the grout is permeable. This could over time allow the penetration of aqueous phases into the waste block and a potential dissolution and release of radionuclides. As a result particularly highly soluble radionuclides may not be retained by the grout. Vitrification could present an alternative, however a similar waste package volume increase may be expected since the encapsulation capacity of glass is potentially similar to or worse than that of grout. FNAG has developed a process for the production of a graphite-glass composite material called Impermeable Graphite Matrix (IGM) [3]. This process is also applicable to irradiated graphite which allows the manufacturing of an impermeable material without volume increase. Crushed i-graphite is mixed with 20 vol.% of glass and then pressed under vacuum at an elevated temperature in an axial hot vacuum press (HVP). The obtained product has zero or negligible porosity and a water impermeable structure. Structural analysis shows that the glass in the composite has replaced the pores in the graphite structure. The typical pore volume of a graphite material is in the range of 20 vol.%. Therefore no volume increase will occur in comparison with the former graphite material. This IGM material will allow the encapsulation of graphite with package densities larger than 1.5 ton per cubic meter. Therefore a huge volume saving can be achieved by such an alternative encapsulation method. Disposal performance is also enhanced since little or no leaching of radionuclides is observed due to the impermeability of the material NNL and FNAG have proved that IGM can be produced by hot isostatic pressing (HIP) which has several advantages for radioactive materials over the HVP process. - The sealed HIP container avoids the release of any radionuclides. - The outside of the waste package is not contaminated. - The HIP process time is shorter than the HVP process time. The isostatic press avoids anisotropic density distributions. - Simple filling of the HIP container has advantages over the filling of an axial die. (authors)« less

Progress of reduction of graphene oxide by ascorbic acid

NASA Astrophysics Data System (ADS)

De Silva, K. Kanishka H.; Huang, Hsin-Hui; Yoshimura, Masamichi

2018-07-01

Graphene oxide (GO) and reduced graphene oxide (RGO) are in greater demand in many research fields. As a result, the synthesis of these materials on a large scale in a costeffective manner is more concerned for numerous applications. In the present work, GO was synthesized by oxidizing natural graphite and reduced by ascorbic acid (AA), which is a green reductant. The reduced products obtained at different time periods were in detail characterized by UV-Visible spectroscopy, X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), attenuated total reflectance Fourier transform infrared (ATR-FT-IR) spectroscopy, Raman spectroscopy, thermogravimetric analysis (TGA), atomic force microscopy (AFM) and scanning electron microscopy (SEM). Results showed that the oxidation of graphite has given highly oxidized GO with a 9.30 Å interlayer space and about 33% of oxygen atomic percentage. Until 50 min of the reduction, both GO and RGO coexist. The reduction rate is fast within the first 30 min. In addition, the suitability of natural graphite over synthetic graphite for the synthesis of GO is shown. The findings of this work pave the way to select GO and RGO for applications of interest in a cheap, green and efficient manner.

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

NASA Astrophysics Data System (ADS)

Peyvandi, Amirpasha

Graphite nanomaterials offer distinct features for effective reinforcement of cementitious matrices in the pre-crack and post-crack ranges of behavior. Thoroughly dispersed and well-bonded nanomaterials provide for effective control of the size and propagation of defects (microcracks) in matrix, and also act as closely spaced barriers against diffusion of moisture and aggressive solutions into concrete. Modified graphite nanomaterials can play multi-faceted roles towards enhancing the mechanical, physical and functional attributes of concrete materials. Graphite nanoplatelets (GP) and carbon nanofibers (CNF) were chosen for use in cementitious materials. Experimental results highlighted the balanced gains in diverse engineering properties of high-performance concrete realized by introduction of graphite nanomaterials. Nuclear Magnetic Resonance (NMR) spectroscopy was used in order to gain further insight into the effects of nanomaterials on the hydration process and structure of cement hydrates. NMR exploits the magnetic properties of certain atomic nuclei, and the sensitivity of these properties to local environments to generate data which enables determination of the internal structure, reaction state, and chemical environment of molecules and bulk materials. 27 Al and 29Si NMR spectroscopy techniques were employed in order to evaluate the effects of graphite nanoplatelets on the structure of cement hydrates, and their resistance to alkali-silica reaction (ASR), chloride ion diffusion, and sulfate attack. Results of 29Si NMR spectroscopy indicated that the percent condensation of C-S-H in cementitious paste was lowered in the presence of nanoplatelets at the same age. The extent of chloride diffusion was assessed indirectly by detecting Friedel's salt as a reaction product of chloride ions with aluminum-bearing cement hydrates. Graphite nanoplatelets were found to significantly reduce the concentration of Friedel's salt at different depths after various periods of exposure to chloride solutions, pointing at the benefits of nanoplatelets towards enhancement of concrete resistance to chloride ion diffusion. It was also found that the intensity of Thaumasite, a key species marking sulfate attack on cement hydrates, was lowered with the addition of graphite nanoplatelets in concrete exposed to sulfate solutions. Experimental evaluations were conducted on scaled-up production of concrete nanocomposite in precast concrete plants. Full-scale reinforced concrete pipes and beams were produced using concrete nanocomposites. Durability and structural tests indicated that the use of graphite nanoplatelets, alone or in combination with synthetic (PVA) fibers, produced significant gains in the durability characteristics, and also benefited the structural performance of precast reinforced concrete products. The material and scaled-up structural investigations conducted in the project concluded that lower-cost graphite nanomaterials (e.g., graphite nanoplatelets) offer significant potentials as multi-functional additives capable of enhancing the barrier, durability and mechanical performance of concrete materials. The benefits of graphite nanomaterials tend to be more pronounced in higher-performance concrete materials.

Analysis of granular flow in a pebble-bed nuclear reactor.

PubMed

Rycroft, Chris H; Grest, Gary S; Landry, James W; Bazant, Martin Z

2006-08-01

Pebble-bed nuclear reactor technology, which is currently being revived around the world, raises fundamental questions about dense granular flow in silos. A typical reactor core is composed of graphite fuel pebbles, which drain very slowly in a continuous refueling process. Pebble flow is poorly understood and not easily accessible to experiments, and yet it has a major impact on reactor physics. To address this problem, we perform full-scale, discrete-element simulations in realistic geometries, with up to 440,000 frictional, viscoelastic 6-cm-diam spheres draining in a cylindrical vessel of diameter 3.5m and height 10 m with bottom funnels angled at 30 degrees or 60 degrees. We also simulate a bidisperse core with a dynamic central column of smaller graphite moderator pebbles and show that little mixing occurs down to a 1:2 diameter ratio. We analyze the mean velocity, diffusion and mixing, local ordering and porosity (from Voronoi volumes), the residence-time distribution, and the effects of wall friction and discuss implications for reactor design and the basic physics of granular flow.

Study of 232Th(n, γ) and 232Th(n,f) reaction rates in a graphite moderated spallation neutron field produced by 1.6 GeV deuterons on lead target

NASA Astrophysics Data System (ADS)

Asquith, N. L.; Hashemi-Nezhad, S. R.; Westmeier, W.; Zhuk, I.; Tyutyunnikov, S.; Adam, J.

2015-02-01

The Gamma-3 assembly of the Joint Institute for Nuclear Research (JINR), Dubna, Russia is designed to emulate the neutron spectrum of a thermal Accelerator Driven System (ADS). It consists of a lead spallation target surrounded by reactor grade graphite. The target was irradiated with 1.6 GeV deuterons from the Nuclotron accelerator and the neutron capture and fission rate of 232Th in several locations within the assembly were experimentally measured. 232Th is a proposed fuel for envisaged Accelerator Driven Systems and these two reactions are fundamental to the performance and feasibility of 232Th in an ADS. The irradiation of the Gamma-3 assembly was also simulated using MCNPX 2.7 with the INCL4 intra-nuclear cascade and ABLA fission/evaporation models. Good agreement between the experimentally measured and calculated reaction rates was found. This serves as a good validation for the computational models and cross section data used to simulate neutron production and transport of spallation neutrons within a thermal ADS.

Briefing Book. Volume 1: The Evolution of the Nuclear Non-Proliferation Regime (Fourth Edition).

DTIC Science & Technology

1998-01-01

usually termed) nuclear reactors. The first of these is that they contain a core or mass of fissile material (the fuel ) which may weigh tens of tons... HTGR is cooled with helium gas and moderated with graphite. Highly enriched uranium is used as fuel (93 per cent U-235), though this may be mixed with...to convert U-238 in a blanket around the core into Pu-239 at a rate faster than its own consumption of fissile material. They thus produce more fuel

Monte-Carlo Simulations of the Nuclear Energy Deposition Inside the CARMEN-1P Differential Calorimeter Irradiated into OSIRIS Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amharrak, H.; Reynard-Carette, C.; Carette, M.

The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. These measurements are then used for other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. Nuclear heating is a great deal of interest at the moment as the measurement of such heating is an important issue for MTRs reactors. This need is especially generated by the new Jules Horowitz Reactor (JHR),more » under construction at CEA/Cadarache 'French Alternative Energies and Atomic Energy Commission'. This new reactor, that will be operational in late 2019, is a new facility for the nuclear research on materials and fuels. Indeed the expected nuclear heating rate is about 20 W/g for nominal capacity of 100 MW. The present Monte Carlo calculation works belong to the IN-CORE (Instrumentation for Nuclear radiation and Calorimetry On line in Reactor): a joint research program between the CEA and Aix- Marseille University in 2009. One scientific aim of this program is to design and develop a multi-sensors device, called CARMEN, dedicated to the measurements of main physical parameters simultaneously encountered inside JHR's experimental channels (core and reflector) such as neutron fluxes, photon fluxes, temperature, and nuclear heating. A first prototype was already developed. This prototype includes two mock-ups dedicated respectively to neutronic measurements (CARMEN-1N) and to photonic measurements (CARMEN-1P) with in particular a specific differential calorimeter. Two irradiation campaigns were performed successfully in the periphery of OSIRIS reactor (a MTR located at Saclay, France) in 2012 for nuclear heating levels up to 2 W/g. First Monte Carlo calculations reduced to the graphite sample of the calorimeter were carried out. A preliminary analysis shows that the numerical results overestimate the measurements by about 20 %. A new approach has been developed in order to estimate the nuclear heating by two methods (energy deposition or KERMA) by considering the whole complete geometry of the sensor. This new approach will contribute to the interpretation of the irradiation campaign and will be useful to improve the out-of-pile calibration procedure of the sensor and its thermal response during irradiations. The aim of this paper is to present simulations made by using MCNP5 Monte-Carlo transport code (using ENDF/B-VI nuclear data library) for the nuclear heating inside the different parts of the calorimeter (head, rod and base). Calculations into two steps will be realized. We will use as an input source in the model new spectra (neutrons, prompt-photons and delayed-photons) calculated with the Monte Carlo code TRIPOLI-4{sup R} inside different experimental channels (water) located into the OSIRIS periphery and used during the CARMEN-1P irradiation campaign. We will consider Neutrons- Photons-Electrons and Photons-Electrons modes. We will begin by a brief description of the differential-calorimeter device geometry. Then the MCNP5 model used for the calculations of nuclear heating inside the calorimeter elements will be introduced. The energy deposition due to the prompt-gamma, delayed-gamma and neutrons, the neutron-activation of the device will be considered. The different components of the nuclear heating inside the different parts of the calorimeter will be detailed. Moreover, a comparison between KERMA and nuclear energy deposition estimations will be given. Finally, a comparison between this total nuclear heating Calculation and Experiment in graphite sample will be determined. (authors)« less

The Hatch-Smolensk exchange

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sproles, A.

1993-03-01

During summer 1992, the World Association of Nuclear Operators (WANO) sponsored an exchange visit between Georgia Power Company's Edwin I. Hatch nuclear plant, a two-unit boiling water reactor site, and the Smolensk atomic energy station, a three-unit RBMK (graphite-moderated and light-water-cooled) plant located 350 km west of Moscow, in Desnogorsk, Russia. The Plant Hatch team included Glenn Goode, manager of engineering support; Curtis Coggin, manager of training and emergency preparedness; Wayne Kirkley, manager of health physics and chemistry; John Lewis, manager of operations; Ray Baker, coordinator of nuclear fuels and contracts; and Bruce McLeod, manager of nuclear maintenance support. Alsomore » traveling with the team was Jerald Towgood, of WANO's Atlanta Centre. The Hatch team visited the Smolensk plant during the week of July 27, 1992.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kalupka, C., E-mail: christian.kalupka@llt.rwth-aachen.de; Finger, J.; Reininghaus, M.

2016-04-21

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

Neutronics and Transient Calculations for the Conversion of the Transient Reactor Rest Facility (TREAT)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Papadias, Dionissios D.

2015-01-01

The Transient Reactor Test Facility (TREAT) is a graphite-reflected, graphitemoderated, and air-cooled reactor fueled with 93.1% enriched UO2 particles dispersed in graphite, with a carbon-to-235U ratio of ~10000:1. TREAT was used to simulate accident conditions by subjecting fuel test samples placed at the center of the core to high energy transient pulses. The transient pulse production is based on the core’s selflimiting nature due to the negative reactivity feedback provided by the fuel graphite as the core temperature rises. The analysis of the conversion of TREAT to low enriched uranium (LEU) is currently underway. This paper presents the analytical methodsmore » used to calculate the transient performance of TREAT in terms of power pulse production and resulting peak core temperatures. The validation of the HEU neutronics TREAT model, the calculation of the temperature distribution and the temperature reactivity feedback as well as the number of fissions generated inside fuel test samples are discussed.« less

Supercritical fluid extraction of bi & multi-layer graphene sheets from graphite by using exfoliation technique

NASA Astrophysics Data System (ADS)

Xavier, Gauravi; Dave, Bhoomi; Khanna, Sakshum

2018-05-01

In recent times, researchers have turned to explore the possibility of using Supercritical Fluid (SCFs) system to penetrate into the inert-gaping of graphite and exfoliate it into a number of layer graphene sheets. The supercritical fluid holds excellent wetting surfaces with low interfacial tension and high diffusion coefficients. Although SCFs exfoliation approach looks promising to developed large scale & low-cost graphene sheet but has not received much attention. To arouse interest and reflection on this approach, this review is organized to summarize the recent progress in graphene production by SCF technology. Here we present the simplest route to obtained layers of graphene sheets by intercalating and exfoliating graphite using supercritical CO2 processing. The layers graphene nano-sheets were collected in dichloromethane (DCM) solution which prevents the restocking of sheets. The obtained graphene sheets show the desired characteristics and thus can be used in physical, chemical and biological sciences. Thus this method provides an effortless and eco-friendly approach for the synthesis of layers of graphene sheets.

Quantification of rare earth elements using laser-induced breakdown spectroscopy

DOE PAGES

Martin, Madhavi; Martin, Rodger C.; Allman, Steve; ...

2015-10-21

In this paper, a study of the optical emission as a function of concentration of laser-ablated yttrium (Y) and of six rare earth elements, europium (Eu), gadolinium (Gd), lanthanum (La), praseodymium (Pr), neodymium (Nd), and samarium (Sm), has been evaluated using the laser-induced breakdown spectroscopy (LIBS) technique. Statistical methodology using multivariate analysis has been used to obtain the sampling errors, coefficient of regression, calibration, and cross-validation of measurements as they relate to the LIBS analysis in graphite-matrix pellets that were doped with elements at several concentrations. Each element (in oxide form) was mixed in the graphite matrix in percentages rangingmore » from 1% to 50% by weight and the LIBS spectra obtained for each composition as well as for pure oxide samples. Finally, a single pellet was mixed with all the elements in equal oxide masses to determine if we can identify the elemental peaks in a mixed pellet. This dataset is relevant for future application to studies of fission product content and distribution in irradiated nuclear fuels. These results demonstrate that LIBS technique is inherently well suited for the future challenge of in situ analysis of nuclear materials. Finally, these studies also show that LIBS spectral analysis using statistical methodology can provide quantitative results and suggest an approach in future to the far more challenging multielemental analysis of ~ 20 primary elements in high-burnup nuclear reactor fuel.« less

Closed tubes preparation of graphite for high-precision AMS radiocarbon analysis

NASA Astrophysics Data System (ADS)

Hajdas, I.; Michczynska, D.; Bonani, G.; Maurer, M.; Wacker, L.

2009-04-01

Radiocarbon dating is an established tool applied in Geochronology. Technical developments of Accelerator Mass Spectrometry AMS, which allow measurements of samples containing less than 1 mg of carbon, opened opportunities for new applications. Moreover, high resolution records of the past changes require high-resolution chronologies i.e. sampling for 14C dating. In result, the field of applications is rapidly expanding and number of radiocarbon analysis is growing rapidly. Nowadays dedicated 14C AMS machines have great capacity for analysis but in order to keep up with the demand for analysis and provide the results as fast as possible a very efficient way of sample preparation is required. Sample preparation for 14C AMS analysis consists of two steps: separation of relevant carbon from the sample material (removing contamination) and preparation of graphite for AMS analysis. The last step usually involves reaction of CO2 with H2, in the presence of metal catalyst (Fe or Co) of specific mesh size heated to 550-625°C, as originally suggested by Vogel et al. (1984). Various graphitization systems have been built in order to fulfil the requirement of sample quality needed for high-precision radiocarbon data. In the early 90ties another method has been proposed (Vogel 1992) and applied by few laboratories mainly for environmental or biomedical samples. This method uses TiH2 as a source of H2 and can be easily and flexibly applied to produce graphite. Sample of CO2 is frozen in to the tube containing pre-conditioned Zn/TiH2 and Fe catalyst. Torch sealed tubes are then placed in the stepwise heated oven at 500/550°C and left to react for several hours. The greatest problem is the lack of control of the reaction completeness and considerable fractionation. However, recently reported results (Xu et al. 2007) suggest that high precision dating using graphite produced in closed tubes might be possible. We will present results of radiocarbon dating of the set of standards and secondary IAEA standards to demonstrate to what level this method can be used for high precision radiocarbon dating. References Vogel JS. 1992. Rapid Production of Graphite without Contamination for Biomedical Ams. Radiocarbon 34: 344-350. Vogel JS, Southon JR, Nelson DE, and Brown TA. 1984. Performance of Catalytically Condensed Carbon for Use in Accelerator Mass-Spectrometry. Nuclear Instruments & Methods in Physics Research Section B-Beam Interactions with Materials and Atoms 233: 289-293. Xu X, Trumbore SE, Zheng S, Southon JR, McDuffee KE, Luttgen M, and Liu JC. 2007. Modifying a sealed tube zinc reduction method for preparation of AMS graphite targets: Reducing background and attaining high precision. Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms Accelerator Mass Spectrometry - Proceedings of the Tenth International Conference on Accelerator Mass Spectrometry 259: 320-329.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aghina, L.O.B.; Bellini, I.W.

The main works and modifications performed on the Argonaut reactor of the Instituto de Engenharia Nuclear after 4 years of operation are described. New positioning and holding system for the fuel elements and graphite wedges for the reactor brought appreciable stability to the operation. The removal of the plastic layer, printing and inspection of the corrosion of the elements plater are also described. (INIS)

Fuselage Structure Response to Boundary Layer, Tonal Sound, and Jet Noise

NASA Technical Reports Server (NTRS)

Maestrello, L.

2004-01-01

Experiments have been conducted to study the response of curved aluminum and graphite-epoxy fuselage structures to flow and sound loads from turbulent boundary layer, tonal sound, and jet noise. Both structures were the same size. The aluminum structure was reinforced with tear stoppers, while the graphite-epoxy structure was not. The graphite-epoxy structure weighed half as much as the aluminum structure. Spatiotemporal intermittence and chaotic behavior of the structural response was observed, as jet noise and tonal sound interacted with the turbulent boundary layer. The fundamental tone distributed energy to other components via wave interaction with the turbulent boundary layer. The added broadband sound from the jet, with or without a shock, influenced the responses over a wider range of frequencies. Instantaneous spatial correlation indicates small localized spatiotemporal regions of convected waves, while uncorrelated patterns dominate the larger portion of the space. By modifying the geometry of the tear stoppers between panels and frame, the transmitted and reflected waves of the aluminum panels were significantly reduced. The response level of the graphite-epoxy structure was higher, but the noise transmitted was nearly equal to that of the aluminum structure. The fundamental shock mode is between 80 deg and 150 deg and the first harmonic is between 20 deg and 80 deg for the underexpanded supersonic jet impinging on the turbulent boundary layer influencing the structural response. The response of the graphite-epoxy structure due to the fundamental mode of the shock impingement was stabilized by an externally fixed oscillator.

Microwave Diffraction Techniques from Macroscopic Crystal Models

ERIC Educational Resources Information Center

Murray, William Henry

1974-01-01

Discusses the construction of a diffractometer table and four microwave models which are built of styrofoam balls with implanted metallic reflecting spheres and designed to simulate the structures of carbon (graphite structure), sodium chloride, tin oxide, and palladium oxide. Included are samples of Bragg patterns and computer-analysis results.…

DC graphite arc furnace, a simple system to reduce mixed waste volume

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wittle, J.K.; Hamilton, R.A.; Trescot, J.

1995-12-31

The volume of low-level radioactive waste can be reduced by the high temperature in a DC Graphite Arc Furnace. This volume reduction can take place with the additional benefit of having the solid residue being stabilized by the vitrified product produced in the process. A DC Graphite Arc Furnace is a simple system in which electricity is used to generate heat to vitrify the material and thermally decompose any organic matter in the waste stream. Examples of this type of waste are protective clothing, resins, and grit blast materials produced in the nuclear industry. The various Department of Energy (DOE)more » complexes produce similar low-level waste streams. Electro-Pyrolysis, Inc. and Svedala/Kennedy Van Saun are engineering and building small 50-kg batch and up to 3,000 kg/hr continuous feed DC furnaces for the remediation, pollution prevention, and decontamination and decommissioning segments of the treatment community. This process has been demonstrated under DOE sponsorship at several facilities and has been shown to produce stable waste forms from surrogate waste materials.« less

Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luther, Erik; Rooyen, Isabella van; Leckie, Rafael

2015-03-01

In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabricationmore » must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.« less

Evidence for Bulk Ripplocations in Layered Solids

NASA Astrophysics Data System (ADS)

Gruber, Jacob; Lang, Andrew C.; Griggs, Justin; Taheri, Mitra L.; Tucker, Garritt J.; Barsoum, Michel W.

2016-09-01

Plastically anisotropic/layered solids are ubiquitous in nature and understanding how they deform is crucial in geology, nuclear engineering, microelectronics, among other fields. Recently, a new defect termed a ripplocation-best described as an atomic scale ripple-was proposed to explain deformation in two-dimensional solids. Herein, we leverage atomistic simulations of graphite to extend the ripplocation idea to bulk layered solids, and confirm that it is essentially a buckling phenomenon. In contrast to dislocations, bulk ripplocations have no Burgers vector and no polarity. In graphite, ripplocations are attracted to other ripplocations, both within the same, and on adjacent layers, the latter resulting in kink boundaries. Furthermore, we present transmission electron microscopy evidence consistent with the existence of bulk ripplocations in Ti3SiC2. Ripplocations are a topological imperative, as they allow atomic layers to glide relative to each other without breaking the in-plane bonds. A more complete understanding of their mechanics and behavior is critically important, and could profoundly influence our current understanding of how graphite, layered silicates, the MAX phases, and many other plastically anisotropic/layered solids, deform and accommodate strain.

Oxidation Behavior of Matrix Graphite and Its Effect on Compressive Strength

DOE PAGES

Zhou, Xiangwen; Contescu, Cristian I.; Zhao, Xi; ...

2017-01-01

Mmore » atrix graphite (G) with incompletely graphitized binder used in high-temperature gas-cooled reactors (HTGRs) is commonly suspected to exhibit lower oxidation resistance in air. In order to reveal the oxidation performance, the oxidation behavior of newly developed A3-3 G at the temperature range from 500 to 950°C in air was studied and the effect of oxidation on the compressive strength of oxidized G specimens was characterized. Results show that temperature has a significant influence on the oxidation behavior of G. The transition temperature between Regimes I and II is ~700°C and the activation energy ( E a ) in Regime I is around 185 kJ/mol, a little lower than that of nuclear graphite, which indicates G is more vulnerable to oxidation. Oxidation at 550°C causes more damage to compressive strength of G than oxidation at 900°C. Comparing with the strength of pristine G specimens, the rate of compressive strength loss is 77.3% after oxidation at 550°C and only 12.5% for oxidation at 900°C. icrostructure images of SE and porosity measurement by ercury Porosimetry indicate that the significant compressive strength loss of G oxidized at 550°C may be attributed to both the uniform pore formation throughout the bulk and the preferential oxidation of the binder.« less

Global analysis of the temperature and flow fields in samples heated in multizone resistance furnaces

NASA Astrophysics Data System (ADS)

Pérez-Grande, I.; Rivas, D.; de Pablo, V.

The temperature field in samples heated in multizone resistance furnaces will be analyzed, using a global model where the temperature fields in the sample, the furnace and the insulation are coupled; the input thermal data is the electric power supplied to the heaters. The radiation heat exchange between the sample and the furnace is formulated analytically, taking into account specular reflections at the sample; for the solid sample the reflectance is both diffuse and specular, and for the melt it is mostly specular. This behavior is modeled through the exchange view factors, which depend on whether the sample is solid or liquid, and, therefore, they are not known a priori. The effect of this specular behavior in the temperature field will be analyzed, by comparing with the case of diffuse samples. A parameter of great importance is the thermal conductivity of the insulation material; it will be shown that the temperature field depends strongly on it. A careful characterization of the insulation is therefore necessary, here it will be done with the aid of experimental results, which will also serve to validate the model. The heating process in the floating-zone technique in microgravity conditions will be simulated; parameters like the Marangoni number or the temperature gradient at the melt-crystal interface will be estimated. Application to the case of compound samples (graphite-silicon-graphite) will be made; the temperature distribution in the silicon part will be studied, especially the temperature difference between the two graphite rods that hold the silicon, since it drives the thermocapillary flow in the melt. This flow will be studied, after coupling the previous model with the convective effects. The possibility of suppresing this flow by the controlled vibration of the graphite rods will be also analyzed. Numerical results show that the thermocapillary flow can indeed be counterbalanced quite effectively.

Indirect measurement of N-14 quadrupolar coupling for NH3 intercalated in potassium graphite

NASA Technical Reports Server (NTRS)

Tsang, T.; Fronko, R. M.; Resing, H. A.

1987-01-01

A method for indirect measurement of the nuclear quadrupolar coupling was developed and applied to NH3 molecules in the graphite intercalation compound K(NH3)4.3C24, which has a layered structure with alternating carbon and intercalant layers. Three triplets were observed in the H-1 NMR spectra of the compound. The value of the N-14 quadrupolar coupling constant of NH3 (3.7 MHz), determined indirectly from the H-1 NMR spectra, was intermediate between the gas value of 4.1 MHz and the solid-state value of 3.2 MHz. The method was also used to deduce the (H-1)-(H-1) and (N-14)-(H-1) dipolar interactions, the H-1 chemical shifts, and the molecular orientations and motions of NH3.

Experimental Studies of Carbon Nanotube Materials for Space Radiators

NASA Technical Reports Server (NTRS)

SanSoucie, MIchael P.; Rogers, Jan R.; Craven, Paul D.; Hyers, Robert W.

2012-01-01

Game ]changing propulsion systems are often enabled by novel designs using advanced materials. Radiator performance dictates power output for nuclear electric propulsion (NEP) systems. Carbon nanotubes (CNT) and carbon fiber materials have the potential to offer significant improvements in thermal conductivity and mass properties. A test apparatus was developed to test advanced radiator designs. This test apparatus uses a resistance heater inside a graphite tube. Metallic tubes can be slipped over the graphite tube to simulate a heat pipe. Several sub ]scale test articles were fabricated using CNT cloth and pitch ]based carbon fibers, which were bonded to a metallic tube using an active braze material. The test articles were heated up to 600 C and an infrared (IR) camera captured the results. The test apparatus and experimental results are presented here.

FY13 Summary Report on the Augmentation of the Spent Fuel Composition Dataset for Nuclear Forensics: SFCOMPO/NF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brady Raap, Michaele C.; Lyons, Jennifer A.; Collins, Brian A.

This report documents the FY13 efforts to enhance a dataset of spent nuclear fuel isotopic composition data for use in developing intrinsic signatures for nuclear forensics. A review and collection of data from the open literature was performed in FY10. In FY11, the Spent Fuel COMPOsition (SFCOMPO) excel-based dataset for nuclear forensics (NF), SFCOMPO/NF was established and measured data for graphite production reactors, Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs) were added to the dataset and expanded to include a consistent set of data simulated by calculations. A test was performed to determine whether the SFCOMPO/NF dataset willmore » be useful for the analysis and identification of reactor types from isotopic ratios observed in interdicted samples.« less

Low Reflection Absorbers for Electromagnetic Waves (Reflexionsarme Absorber Fuer Elektromagnetische Wellen)

DTIC Science & Technology

1960-11-01

should be briefly recalled. They consist of air sound, as a rule, of porous substances, mineral wool , glass wool, and similar substances, whose...nonreflecting room, then small particles of graphite will -17- pi m u . I be inserted into the pores of the porous glass wool or mineral wool . Such wedges

Observation of Persistent Currents in Finely Dispersed Pyrolytic Graphite

NASA Astrophysics Data System (ADS)

Saad, M.; Gilmutdinov, I. F.; Kiiamov, A. G.; Tayurskii, D. A.; Nikitin, S. I.; Yusupov, R. V.

2018-01-01

The trapped magnetic flux in the finely ground pyrolytic graphite sample annealed at 670 K in air has been observed. Flux trapping occurs on cooling of the sample from room temperature to 10 K in a magnetic field of 1 T. The magnitude and sign of the induced trapped moment remain unchanged when the applied magnetic field is varied within ±1 T at T K. The trapped magnetic flux is manifested in the displacement of the magnetization curve relative to that of the sample cooled in zero field. Displacement magnitude gradually decreases with the temperature increase up to 350 K, not reaching zero. The set of experimental observations probably reflects the presence in the sample of a granular high-temperature superconducting phase.

Nuclear Data Needs for Generation IV Nuclear Energy Systems

NASA Astrophysics Data System (ADS)

Rullhusen, Peter

2006-04-01

Nuclear data needs for generation IV systems. Future of nuclear energy and the role of nuclear data / P. Finck. Nuclear data needs for generation IV nuclear energy systems-summary of U.S. workshop / T. A. Taiwo, H. S. Khalil. Nuclear data needs for the assessment of gen. IV systems / G. Rimpault. Nuclear data needs for generation IV-lessons from benchmarks / S. C. van der Marck, A. Hogenbirk, M. C. Duijvestijn. Core design issues of the supercritical water fast reactor / M. Mori ... [et al.]. GFR core neutronics studies at CEA / J. C. Bosq ... [et al]. Comparative study on different phonon frequency spectra of graphite in GCR / Young-Sik Cho ... [et al.]. Innovative fuel types for minor actinides transmutation / D. Haas, A. Fernandez, J. Somers. The importance of nuclear data in modeling and designing generation IV fast reactors / K. D. Weaver. The GIF and Mexico-"everything is possible" / C. Arrenondo Sánchez -- Benmarks, sensitivity calculations, uncertainties. Sensitivity of advanced reactor and fuel cycle performance parameters to nuclear data uncertainties / G. Aliberti ... [et al.]. Sensitivity and uncertainty study for thermal molten salt reactors / A. Biduad ... [et al.]. Integral reactor physics benchmarks- The International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPHEP) / J. B. Briggs, D. W. Nigg, E. Sartori. Computer model of an error propagation through micro-campaign of fast neutron gas cooled nuclear reactor / E. Ivanov. Combining differential and integral experiments on [symbol] for reducing uncertainties in nuclear data applications / T. Kawano ... [et al.]. Sensitivity of activation cross sections of the Hafnium, Tanatalum and Tungsten stable isotopes to nuclear reaction mechanisms / V. Avrigeanu ... [et al.]. Generating covariance data with nuclear models / A. J. Koning. Sensitivity of Candu-SCWR reactors physics calculations to nuclear data files / K. S. Kozier, G. R. Dyck. The lead cooled fast reactor benchmark BREST-300: analysis with sensitivity method / V. Smirnov ... [et al.]. Sensitivity analysis of neutron cross-sections considered for design and safety studies of LFR and SFR generation IV systems / K. Tucek, J. Carlsson, H. Wider -- Experiments. INL capabilities for nuclear data measurements using the Argonne intense pulsed neutron source facility / J. D. Cole ... [et al.]. Cross-section measurements in the fast neutron energy range / A. Plompen. Recent measurements of neutron capture cross sections for minor actinides by a JNC and Kyoto University Group / H. Harada ... [et al.]. Determination of minor actinides fission cross sections by means of transfer reactions / M. Aiche ... [et al.] -- Evaluated data libraries. Nuclear data services from the NEA / H. Henriksson, Y. Rugama. Nuclear databases for energy applications: an IAEA perspective / R. Capote Noy, A. L. Nichols, A. Trkov. Nuclear data evaluation for generation IV / G. Noguère ... [et al.]. Improved evaluations of neutron-induced reactions on americium isotopes / P. Talou ... [et al.]. Using improved ENDF-based nuclear data for candu reactor calculations / J. Prodea. A comparative study on the graphite-moderated reactors using different evaluated nuclear data / Do Heon Kim ... [et al.].

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

NASA Astrophysics Data System (ADS)

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 grades are not large, the results of the oxidation and hardness test show a higher irradiation sensitivity for the IG-110. The similar irradiation sensitivities between the grades were attributed to the same degree of graphitization (DOG) of the grades.

Three-dimensional NDE of VHTR core components via simulation-based testing. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guzina, Bojan; Kunerth, Dennis

2014-09-30

A next generation, simulation-driven-and-enabled testing platform is developed for the 3D detection and characterization of defects and damage in nuclear graphite and composite structures in Very High Temperature Reactors (VHTRs). The proposed work addresses the critical need for the development of high-fidelity Non-Destructive Examination (NDE) technologies for as-manufactured and replaceable in-service VHTR components. Centered around the novel use of elastic (sonic and ultrasonic) waves, this project deploys a robust, non-iterative inverse solution for the 3D defect reconstruction together with a non-contact, laser-based approach to the measurement of experimental waveforms in VHTR core components. In particular, this research (1) deploys three-dimensionalmore » Scanning Laser Doppler Vibrometry (3D SLDV) as a means to accurately and remotely measure 3D displacement waveforms over the accessible surface of a VHTR core component excited by mechanical vibratory source; (2) implements a powerful new inverse technique, based on the concept of Topological Sensitivity (TS), for non-iterative elastic waveform tomography of internal defects - that permits robust 3D detection, reconstruction and characterization of discrete damage (e.g. holes and fractures) in nuclear graphite from limited-aperture NDE measurements; (3) implements state-of-the art computational (finite element) model that caters for accurately simulating elastic wave propagation in 3D blocks of nuclear graphite; (4) integrates the SLDV testing methodology with the TS imaging algorithm into a non-contact, high-fidelity NDE platform for the 3D reconstruction and characterization of defects and damage in VHTR core components; and (5) applies the proposed methodology to VHTR core component samples (both two- and three-dimensional) with a priori induced, discrete damage in the form of holes and fractures. Overall, the newly established SLDV-TS testing platform represents a next-generation NDE tool that surpasses all existing techniques for the 3D ultrasonic imaging of material damage from non-contact, limited-aperture waveform measurements. Outlook. The next stage in the development of this technology includes items such as (a) non-contact generation of mechanical vibrations in VHTR components via thermal expansion created by high-intensity laser; (b) development and incorporation of Synthetic Aperture Focusing Technique (SAFT) for elevating the accuracy of 3D imaging in highly noisy environments with minimal accessible surface; (c) further analytical and computational developments to facilitate the reconstruction of diffuse damage (e.g. microcracks) in nuclear graphite as they lead to the dispersion of elastic waves, (d) concept of model updating for accurate tracking of the evolution of material damage via periodic inspections; (d) adoption of the Bayesian framework to obtain information on the certainty of obtained images; and (e) optimization of the computational scheme toward real-time, model-based imaging of damage in VHTR core components.« less

Corrosion-induced microstructural developments in 316 stainless steel during exposure to molten Li2BeF4(FLiBe) salt

NASA Astrophysics Data System (ADS)

Zheng, Guiqiu; He, Lingfeng; Carpenter, David; Sridharan, Kumar

2016-12-01

The microstructural developments in the near-surface regions of AISI 316 stainless steel during exposure to molten Li2BeF4 (FLiBe) salt have been investigated with the goal of using this material for the construction of the fluoride salt-cooled high-temperature reactor (FHR), a leading nuclear reactor concept for the next generation nuclear plants (NGNP). Tests were conducted in molten FLiBe salt (melting point: 459 °C) at 700 °C in graphite crucibles and 316 stainless steel crucibles for exposure duration of up to 3000 h. Corrosion-induced microstructural changes in the near-surface regions of the samples were characterized using scanning electron microscopy (SEM) in conjunction with energy dispersive x-ray spectroscopy (EDS) and electron backscatter diffraction (EBSD), and scanning transmission electron microscopy (STEM) with EDS capabilities. Intergranular corrosion attack in the near-surface regions was observed with associated Cr depletion along the grain boundaries. High-angle grain boundaries (15-180°) were particularly prone to intergranular attack and Cr depletion. The depth of attack extended to the depths of 22 μm after 3000-h exposure for the samples tested in graphite crucible, while similar exposure in 316 stainless steel crucible led to the attack depths of only about 11 μm. Testing in graphite crucibles led to the formation of nanometer-scale Mo2C, Cr7C3 and Al4C3 particle phases in the near-surface regions of the material. The copious depletion of Cr in the near-surface regions induced a γ-martensite to α-ferrite phase (FeNix) transformation. Based on the microstructural analysis, a thermal diffusion controlled corrosion model was developed and experimentally validated for predicting long-term corrosion attack depth.

Carbon chemistry: The high temperature syntheses and applications of nanotubes andsp-hybridized compounds

NASA Astrophysics Data System (ADS)

Mitchell, Daniel Robert

A brief introduction to carbon chemistry is given with an emphasis on the use high-temperature reactions that use carbon vapor, generated from graphite, to synthesize nano-structured materials. Laser and electric are ablation of graphite was utilized to create a variety of high carbon content materials ranging from discrete acetylenic molecules to extremely large multi-wall nanotubes. A new synthesis for large carbon nanotubes, containing 1--5 atom percent nitrogen bound into the graphite lattice, was realized by the reaction of carbon vapor, nickel/yttrium catalyst and cyanogen gas. These carbon "megatubes" were then employed as a substrate to tether a wide variety of molecules both inorganic and organic. The megatubes, in their native and derivatized states, were then assembled into simple circuits to explore their electronic transport properties. Direct fluorination was used to post-treat the surface of the multi-wall carbon nanotubes in order to alter the inherent physical and chemical properties of the tubes, as well as to serve as another route to functionalize their surfaces. Fluorine sites on the walls of the tube were allowed to react with Grignard reagents to produce nantoubes with the chosen alkyl chemically bonded to the surface. Products were characterized with techniques similar to unfluorinated tubules. Using similar carbon vaporization techniques, sp-hybridized carbon chain compounds were synthesized. Using a one-step method dicyanopolyynes were synthesized and characterized with nuclear magnetic resonance and mass spectroscopy, containing up to 8 acetylenic repeat units. A two-step method was also utilized to create polyynes terminated with trifluoromethyl or nitrile radicals generated in a capacitively coupled radio frequency glow plasma discharge. A partial characterization of these products was accomplished with nuclear magnetic resonance, mass, and infrared spectroscopy techniques.

Two-dimensional network stability of nucleobases and amino acids on graphite under ambient conditions: adenine, L-serine and L-tyrosine.

PubMed

Bald, Ilko; Weigelt, Sigrid; Ma, Xiaojing; Xie, Pengyang; Subramani, Ramesh; Dong, Mingdong; Wang, Chen; Mamdouh, Wael; Wang, Jianguo; Besenbacher, Flemming

2010-04-14

We have investigated the stability of two-dimensional self-assembled molecular networks formed upon co-adsorption of the DNA base, adenine, with each of the amino acids, L-serine and L-tyrosine, on a highly oriented pyrolytic graphite (HOPG) surface by drop-casting from a water solution. L-serine and L-tyrosine were chosen as model systems due to their different interaction with the solvent molecules and the graphite substrate, which is reflected in a high and low solubility in water, respectively, compared with adenine. Combined scanning tunneling microscopy (STM) measurements and density functional theory (DFT) calculations show that the self-assembly process is mainly driven by the formation of strong adenine-adenine hydrogen bonds. We find that pure adenine networks are energetically more stable than networks built up of either pure L-serine, pure L-tyrosine or combinations of adenine with L-serine or L-tyrosine, and that only pure adenine networks are stable enough to be observable by STM under ambient conditions.

Comparative Studies on UO2 Fueled HTTR Several Nuclear Data Libraries

NASA Astrophysics Data System (ADS)

Hidayati, Anni N.; Prastyo, Puguh A.; Waris, Abdul; Irwanto, Dwi

2017-07-01

HTTR (High Temperature Engineering Test Reactor) is one of Generation IV nuclear reactors that has been developed by JAERI (former name of JAEA, JAPAN). HTTR uses graphite moderator, helium gas coolant with UO2 fuel and outlet coolant temperature of 900°C or higher than that. Several studies regarding HTTR have been performed by employing JENDL 3.2 nuclear data libraries. In this paper, comparative evaluation of HTTR with several nuclear data libraries (JENDL 3.3, JENDL 4.0, and JEF 3.1) have been conducted.. The 3-D calculation was performed by using CITATION module of SRAC 2006 code. The result shows some differences between those nuclear data libraries result. K-eff or core effective multiplication factor results are about 1.17, 1,18 and 1,19 (JENDL 3.3, JENDL 4.0, and JEF 3.1) at Begin of Life, also at the End of Life (after two years operation) are 1.16, 1.17 and 1.17 for each nuclear data libraries. There are some different result of K-eff but for neutron spectra results, those nuclear data libraries show the same result.

Spectral reflectance properties of carbon-bearing materials

NASA Technical Reports Server (NTRS)

Cloutis, Edward A.; Gaffey, Michael J.; Moslow, Thomas F.

1994-01-01

The 0.3-2.6 micrometers spectral reflectance properties of carbon polymorphs (graphite, carbon black, diamond), carbides (silicon carbide, cementite), and macromolecular organic-bearing materials (coal, coal tar extract, oil sand, oil shale) are found to vary from sample to sample and among groups. The carbon polymorphs are readily distinguishable on the basis of their visible-near infrared spectral slopes and shapes. The spectra of macromolecular organic-bearing materials show increases in reflectance toward longer wavelengths, exceeding the reflectance rise of more carbon-rich materials. Reflectance spectra of carbonaceous materials are affected by the crystal structure, composition, and degree of order/disorder of the samples. The characteristic spectral properties can potentially be exploited to identify individual carbonaceous grains in meteorites (as separates or in situ) or to conduct remote sensing geothermometry and identification of carbonaceous phases on asteroids.

Calculated criticality for sup 235 U/graphite systems using the VIM Monte Carlo code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collins, P.J.; Grasseschi, G.L.; Olsen, D.N.

1992-01-01

Calculations for highly enriched uranium and graphite systems gained renewed interest recently for the new production modular high-temperature gas-cooled reactor (MHTGR). Experiments to validate the physics calculations for these systems are being prepared for the Transient Reactor Test Facility (TREAT) reactor at Argonne National Laboratory (ANL-West) and in the Compact Nuclear Power Source facility at Los Alamos National Laboratory. The continuous-energy Monte Carlo code VIM, or equivalently the MCNP code, can utilize fully detailed models of the MHTGR and serve as benchmarks for the approximate multigroup methods necessary in full reactor calculations. Validation of these codes and their associated nuclearmore » data did not exist for highly enriched {sup 235}U/graphite systems. Experimental data, used in development of more approximate methods, dates back to the 1960s. The authors have selected two independent sets of experiments for calculation with the VIM code. The carbon-to-uranium (C/U) ratios encompass the range of 2,000, representative of the new production MHTGR, to the ratio of 10,000 in the fuel of TREAT. Calculations used the ENDF/B-V data.« less

Experimental Study and Computational Simulations of Key Pebble Bed Thermo-mechanics Issues for Design and Safety

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tokuhiro, Akira; Potirniche, Gabriel; Cogliati, Joshua

2014-07-08

An experimental and computational study, consisting of modeling and simulation (M&S), of key thermal-mechanical issues affecting the design and safety of pebble-bed (PB) reactors was conducted. The objective was to broaden understanding and experimentally validate thermal-mechanic phenomena of nuclear grade graphite, specifically, spheres in frictional contact as anticipated in the bed under reactor relevant pressures and temperatures. The contact generates graphite dust particulates that can subsequently be transported into the flowing gaseous coolent. Under postulated depressurization transients and with the potential for leaked fission products to be adsorbed onto graphite 'dust', there is the potential for fission products to escapemore » from the primary volume. This is a design safety concern. Furthermore, earlier safety assessment identified the distinct possibility for the dispersed dust to combust in contact with air if sufficient conditions are met. Both of these phenomena were noted as important to design review and containing uncertainty to warrant study. The team designed and conducted two separate effects tests to study and benchmark the potential dust-generation rate, as well as study the conditions under which a dust explosion may occure in a standardized, instrumented explosion chamber.« less

Nuclear reactor control

DOEpatents

Cawley, William E.; Warnick, Robert F.

1982-01-01

1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

Graphene levitation and orientation control using a magnetic field

NASA Astrophysics Data System (ADS)

Niu, Chao; Lin, Feng; Wang, Zhiming M.; Bao, Jiming; Hu, Jonathan

2018-01-01

This paper studies graphene levitation and orientation control using a magnetic field. The torques in all three spatial directions induced by diamagnetic forces are used to predict stable conditions for different shapes of millimeter-sized graphite plates. We find that graphite plates, in regular polygon shapes with an even number of sides, will be levitated in a stable manner above four interleaved permanent magnets. In addition, the orientation of micrometer-sized graphene flakes near a permanent magnet is studied in both air and liquid environments. Using these analyses, we are able to simulate optical transmission and reflection on a writing board and thereby reveal potential applications using this technology for display screens. Understanding the control of graphene flake orientation will lead to the discovery of future applications using graphene flakes.

Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bruce G. Schnitzler; Stanley K. Borowski

Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified asmore » the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were constrained to fit within the payload volume of the then planned space shuttle. The SNRE core design utilized hexagonal fuel elements and hexagonal structural support elements. The total number of elements can be varied to achieve engine designs of higher or lower thrust levels. Some variation in the ratio of fuel elements to structural elements is also possible. Options for SNRE-based engine designs in the 25,000-lbf thrust range were described in a recent (2010) Joint Propulsion Conference paper. The reported designs met or exceeded the performance characteristics baselined in the DRA 5.0 Study. Lower thrust SNRE-based designs were also described in a recent (2011) Joint Propulsion Conference paper. Recent activities have included parallel evaluation and design efforts on fast spectrum engines employing refractory metal alloy fuels. These efforts include evaluation of both heritage designs from the Argonne National Laboratory (ANL) and General Electric Company GE-710 Programs as well as more recent designs. Results are presented for a number of not-yet optimized fast spectrum engine options.« less

Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Schnitzler, Bruce G.; Borowski, Stanley K.

2012-01-01

Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were constrained to fit within the payload volume of the then planned space shuttle. The SNRE core design utilized hexagonal fuel elements and hexagonal structural support elements. The total number of elements can be varied to achieve engine designs of higher or lower thrust levels. Some variation in the ratio of fuel elements to structural elements is also possible. Options for SNRE-based engine designs in the 25,000-lbf thrust range were described in a recent (2010) Joint Propulsion Conference paper. The reported designs met or exceeded the performance characteristics baselined in the DRA 5.0 Study. Lower thrust SNRE-based designs were also described in a recent (2011) Joint Propulsion Conference paper. Recent activities have included parallel evaluation and design efforts on fast spectrum engines employing refractory metal alloy fuels. These efforts include evaluation of both heritage designs from the Argonne National Laboratory (ANL) and General Electric Company GE-710 Programs as well as more recent designs. Results are presented for a number of not-yet optimized fast spectrum engine options.

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

NASA Astrophysics Data System (ADS)

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

2018-02-01

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

Comparison of W-VC-C composites against Co-60, Se-75 and Sb-125 for gamma radioisotope sources

NASA Astrophysics Data System (ADS)

Demir, Ertugrul; Tugrul, A. Beril; Buyuk, Bulent; Yilmaz, Ozan; Ovecoglu, Lutfi

2018-02-01

Tungsten based materials are considered to be the promising materials for nuclear applications due to the good properties. The tungsten composite materials have so many advantages in nuclear technological applications especially fusion reactor systems. In this paper, Tungsten-Vanadium carbide-Graphite (W-VC-C) which include 93% tungsten (W), 6% vanadium carbide (VC) and 1% graphite (C) also which has three different alloying time (6-12-24 hours) were produced by mechanical alloying method. Co-60, Se-75 and Sb-125 gamma radioisotopeswere used as a gamma sources in order to determine behavior of gamma attenuation properties of the composite materials. The experimental results were compared with each other to clarify effects of varying gamma energies on the tungsten based composite materials. The mass attenuation coefficients of the samples were obtained by using XCOM computer code and compared with experimental data. The gamma linear attenuation, the mass attenuation coefficients and half value thickness (HVL) of the samples were evaluated and compared with Co-60, Se-75 and Sb-125 for gamma radioisotopes. Results showed that gamma attenuation coefficients of the samples depend on gamma energies and mechanical alloying time has negatively effect on the gamma shielding properties for the all studied W-VC-C.

Application of X-ray microcomputed tomography in the characterization of irradiated nuclear fuel and material specimens

DOE PAGES

Silva, Chinthaka M.; Snead, Lance Lewis; Hunn, John D.; ...

2015-08-03

X-ray microcomputed tomography (µCT) was applied in characterizing the internal structures of a number of irradiated materials, including carbon-carbon fibre composites, nuclear-grade graphite and tristructural isotropic-coated fuel particles. Local cracks in carbon-carbon fibre composites associated with their synthesis process were observed with µCT without any destructive sample preparation. Pore analysis of graphite samples was performed quantitatively, and qualitative analysis of pore distribution was accomplished. It was also shown that high-resolution µCT can be used to probe internal layer defects of tristructural isotropic-coated fuel particles to elucidate the resulting high release of radioisotopes. Layer defects of sizes ranging from 1 tomore » 5 µm and up could be isolated by to-mography. As an added advantage, µCT could also be used to identify regions with high densities of radioisotopes to deter-mine the proper plane and orientation of particle mounting for further analytical characterization, such as materialographic sectioning followed by optical and electron microscopy. Lastly, in fully ceramic matrix fuel forms, despite the highly absorbing matrix, characterization of tristructural isotropic-coated particles embedded in a silicon carbide matrix was accomplished usingµCT and related advanced image analysis techniques.« less

Investigation of Isotopically Tailored Boron in Advanced Fission and Fusion Reactor Systems.

NASA Astrophysics Data System (ADS)

Domaszek, Gerald Raymond

This research examines the use of B^ {11}, in the form of metallic boron and boron carbide, as a moderating and reflecting material. An examination of the neutronic characteristics of the B ^{11} isotope of boron has revealed that B^{11} has neutron scattering and absorption cross sections favorably comparable to those of Be^9 and C^ {12}. Preliminary analysis of the neutronics of B ^{11} were performed by conducting one dimensional transport calculations on an infinite slab of varying thickness. Beryllium is the best of the three materials in reflecting neutrons due primarily to the contribution from (n,2n) reactions. Tailored neutron energy beam transmission experiments were carried out to experimentally verify the predicted neutronic characteristics of B^{11 }. To further examine the neutron moderating and reflecting characteristics of B^{11 }, the energy dependent neutron flux was measured as a function of position in an exponential pile constructed of B_4C isotopically enriched to 98.5 percent B^{11}. After the experimental verification of the neutronic behavior of B^{11}, further design studies were conducted using metallic boron and boron carbide enriched in the B^{11 } isotope. The use of materials isotopically enriched in B^{11} as a liner in the first wall/blanket of a magnetic confinement fusion reactor demonstrated acceptable tritium regeneration in the lithium blanket. Analysis of the effect of contaminant levels of B^{10} showed that B^{10} contents of less than 1 percent in metallic boron produced negligible adverse effects on the tritium breeding. A comparison of the effectiveness of graphite and B^{11}_4C when used as moderators in a reactor fueled with natural uranium has shown that the maximum k_infty for a given fuel rod design is approximately the same for both materials. Approximately half the volume of the moderator is required when B^{11 }_4C is substituted for graphite to obtain essentially the same K_infty . An analysis of the effectiveness of various materials as reflector control elements for a compact space reactor has shown that B^{11} is neutronically superior to graphite in these applications. Metallic boron and boron carbide isotopically enriched in B^{11} have been demonstrated to be neutronically acceptable for varied applications in advanced reactor systems. B^ {11} has been shown to be superior in performance to graphite. While only somewhat inferior to beryllium as neutron multipliers, B^ {11} and B^{11} _4C have safety, supply and cost advantage over beryllium. (Abstract shortened with permission of author.).

Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

NASA Astrophysics Data System (ADS)

Cisneros, Anselmo Tomas, Jr.

The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP and PEBBED for a high temperature gas cooled pebble bed reactor. Three parametric studies were performed for exploring the design space of the PB-FHR---to select a fuel design for the PB-FHR] to select a core configuration; and to optimize the PB-FHR design. These parametric studies investigated trends in the dependence of important reactor performance parameters such as burnup, temperature reactivity feedback, radiation damage, etc on the reactor design variables and attempted to understand the underlying reactor physics responsible for these trends. A pebble fuel parametric study determined that pebble fuel should be designed with a carbon to heavy metal ratio (C/HM) less than 400 to maintain negative coolant temperature reactivity coefficients. Seed and thorium blanket-, seed and inert pebble reflector- and seed only core configurations were investigated for annular FHR PBRs---the C/HM of the blanket pebbles and discharge burnup of the thorium blanket pebbles were additional design variable for core configurations with thorium blankets. Either a thorium blanket or graphite pebble reflector is required to shield the outer graphite reflector enough to extend its service lifetime to 60 EFPY. The fuel fabrication costs and long cycle lengths of the thorium blanket fuel limit the potential economic advantages of using a thorium blanket. Therefore, the seed and pebble reflector core configuration was adopted as the baseline core configuration. Multi-objective optimization with respect to economics was performed for the PB-FHR accounting for safety and other physical design constraints derived from the high-level safety regulatory criteria. These physical constraints were applied along in a design tool, Nuclear Application Value Estimator, that evaluated a simplified cash flow economics model based on estimates of reactor performance parameters calculated using correlations based on the results of parametric design studies for a specific PB-FHR design and a set of economic assumptions about the electricity market to evaluate the economic implications of design decisions. The optimal PB-FHR design---Mark 1 PB-FHR---is described along with a detailed summary of its performance characteristics including: the burnup, the burnup evolution, temperature reactivity coefficients, the power distribution, radiation damage distributions, control element worths, decay heat curves and tritium production rates. The Mk1 PB-FHR satisfies the PB-FHR safety criteria. The fuel, moderator (pebble core, pebble shell, graphite matrix, TRISO layers) and coolant have global negative temperature reactivity coefficients and the fuel temperatures are well within their limits.

Evaluation of Cadmium Ratio and Foil Activation Measurements for a Beryllium-Reflected Assembly of U(93.15)O 2 Fuel Rods (1.506-cm Triangular Pitch)

DOE PAGES

Marshall, Margaret A.

2014-11-04

A series of small, compact critical assembly (SCCA) experiments were completed from 1962 to 1965 at Oak Ridge National Laboratory’s Critical Experiments Facility (ORCEF) in support of the Medium-Power Reactor Experiments (MPRE) program. Initial experiments, performed in November and December of 1962, consisted of a core of un-moderated stainless-steel tubes, each containing 26 UOIdaho National Laboratory (INL), Idaho Falls, ID (United States) fuel pellets, surrounded by a graphite reflector. Measurements were performed to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. The graphite reflectors were then changed to beryllium reflectors. For the beryllium reflected assemblies, the fuel wasmore » in 1.506-cm-triangular and 7-tube clusters leading to two critical configurations. Once the critical configurations had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U, and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements, performed on the 1.506-cm-array critical configuration, have been evaluated and are described in this paper.« less

Measurement of Thick Target Neutron Yields at 0-Degree Bombarded With 140-MeV, 250-MeV And 350-MeV Protons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Iwamoto, Yosuke; /JAERI, Kyoto; Taniguchi, Shingo

Neutron energy spectra at 0{sup o} produced from stopping-length graphite, aluminum, iron and lead targets bombarded with 140, 250 and 350 MeV protons were measured at the neutron TOF course in RCNP of Osaka University. The neutron energy spectra were obtained by using the time-of-flight technique in the energy range from 10 MeV to incident proton energy. To compare the experimental results, Monte Carlo calculations with the PHITS and MCNPX codes were performed using the JENDL-HE and the LA150 evaluated nuclear data files, the ISOBAR model implemented in PHITS, and the LAHET code in MCNPX. It was found that thesemore » calculated results at 0{sup o} generally agreed with the experimental results in the energy range above 20 MeV except for graphite at 250 and 350 MeV.« less

Comparison of graphite, aluminum, and TransHab shielding material characteristics in a high-energy neutron field

NASA Technical Reports Server (NTRS)

Badhwar, G. D.; Huff, H.; Wilkins, R.; Thibeault, Sheila

2002-01-01

Space radiation transport models clearly show that low atomic weight materials provide a better shielding protection for interplanetary human missions than high atomic weight materials. These model studies have concentrated on shielding properties against charged particles. A light-weight, inflatable habitat module called TransHab was built and shown to provide adequate protection against micrometeoroid impacts and good shielding properties against charged particle radiation in the International Space Station orbits. An experiment using a tissue equivalent proportional counter, to study the changes in dose and lineal energy spectra with graphite, aluminum, and a TransHab build-up as shielding, was carried out at the Los Alamos Nuclear Science Center neutron facility. It is a continuation of a previous study using regolith and doped polyethylene materials. This paper describes the results and their comparison with the previous study. Published by Elsevier Science Ltd.

Student research with 400keV beams: {sup 13}N radioisotope production target development

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fru, L. Che; Clymer, J.; Compton, N.

2013-04-19

The AN400 Van de Graaff accelerator at the Minnesota State University, Mankato, Applied Nuclear Science Lab has demonstrated utility as an accessible and versatile platform for student research. Despite the limits of low energy, the research team successfully developed projects with applications to the wider radioisotope production community. A target system has been developed for producing and extracting {sup 13}N by the {sup 12}C(d,n){sup 13}N reaction below 400keV. The system is both reusable and robust, with future applications to higher energy machines producing this important radioisotope for physiological imaging studies with Positron Emission Tomography. Up to 36({+-}1)% of the {supmore » 13}N was extracted from the graphite matrix when 35 A current was externally applied to the graphite target while simultaneously flushing the target chamber with CO{sub 2} gas.« less

Advanced Thin Ionization Calorimeter (ATIC) Update

NASA Technical Reports Server (NTRS)

Ahn, H. S.; Ganel, O.; Kim, K. C.; Seo, E. S.; Sina, R.; Wang, J. Z.; Wu, J.; Case, G.; Ellison, S. B.; Gould, R.;

Graphite fiber/copper matrix composites for space power heat pipe fin applications

NASA Astrophysics Data System (ADS)

McDanels, David L.; Baker, Karl W.; Ellis, David L.

1991-01-01

High specific thermal conductivity (thermal conductivity divided by density) is a major design criterion for minimizing system mass for space power systems. For nuclear source power systems, graphite fiber reinforced copper matrix (Gr/Cu) composites offer good potential as a radiator fin material operating at service temperatures above 500 K. Specific thermal conductivity in the longitudinal direction is better than beryllium and almost twice that of copper. The high specific thermal conductivity of Gr/Cu offers the potential of reducing radiator mass by as much as 30 percent. Gr/Cu composites also offer the designer a range of available properties for various missions and applications. The properties of Gr/Cu are highly anisotropic. Longitudinal elastic modulus is comparable to beryllium and about three times that of copper. Thermal expansion in the longitudinal direction is near zero, while it exceeds that of copper in the transverse direction.

Integrated Solar Upper Stage Technical Support

NASA Technical Reports Server (NTRS)

Jaworske, Donald A.

1998-01-01

NASA Lewis Research Center is participating in the Integrated Solar Upper Stage (ISUS) program. This program is a ground-based demonstration of an upper stage concept that will be used to generate both solar propulsion and solar power. Solar energy collected by a primary concentrator is directed into the aperture of a secondary concentrator and further concentrated into the aperture of a heat receiver. The energy stored in the receiver-absorber-converter is used to heat hydrogen gas to provide propulsion during the orbital transfer portion of the mission. During the balance of the mission, electric power is generated by thermionic diodes. Several materials issues were addressed as part of the technical support portion of the ISUS program, including: 1) Evaluation of primary concentrator coupons; 2) Evaluation of secondary concentrator coupons; 3) Evaluation of receiver-absorber-converter coupons; 4) Evaluation of in-test witness coupons. Two different types of primary concentrator coupons were evaluated from two different contractors-replicated coupons made from graphite-epoxy composite and coupons made from microsheet glass. Specular reflectivity measurements identified the replicated graphite-epoxy composite coupons as the primary concentrator material of choice. Several different secondary concentrator materials were evaluated, including a variety of silver and rhodium reflectors. The specular reflectivity of these materials was evaluated under vacuum at temperatures up to 800 C. The optical properties of several coupons of rhenium on graphite were evaluated to predict the thermal performance of the receiver-absorber-converter. Finally, during the ground test demonstration, witness coupons placed in strategic locations throughout the thermal vacuum facility were evaluated for contaminants. All testing for the ISUS program was completed successfully in 1997. Investigations related to materials issues have proven helpful in understanding the operation of the test article, leading to a potential ISUS flight test in 2002.

Nuclear Science Symposium, 21st, Scintillation and Semiconductor Counter Symposium, 14th, and Nuclear Power Systems Symposium, 6th, Washington, D.C., December 11-13, 1974, Proceedings

NASA Technical Reports Server (NTRS)

1975-01-01

Papers are presented dealing with latest advances in the design of scintillation counters, semiconductor radiation detectors, gas and position sensitive radiation detectors, and the application of these detectors in biomedicine, satellite instrumentation, and environmental and reactor instrumentation. Some of the topics covered include entopistic scintillators, neutron spectrometry by diamond detector for nuclear radiation, the spherical drift chamber for X-ray imaging applications, CdTe detectors in radioimmunoassay analysis, CAMAC and NIM systems in the space program, a closed loop threshold calibrator for pulse height discriminators, an oriented graphite X-ray diffraction telescope, design of a continuous digital-output environmental radon monitor, and the optimization of nanosecond fission ion chambers for reactor physics. Individual items are announced in this issue.

The nuclear battery

NASA Astrophysics Data System (ADS)

Kozier, K. S.; Rosinger, H. E.

The evolution and present status of an Atomic Energy of Canada Limited program to develop a small, solid-state, passively cooled reactor power supply known as the Nuclear Battery is reviewed. Key technical features of the Nuclear Battery reactor core include a heat-pipe primary heat transport system, graphite neutron moderator, low-enriched uranium TRISO coated-particle fuel and the use of burnable poisons for long-term reactivity control. An external secondary heat transport system extracts useful heat energy, which may be converted into electricity in an organic Rankine cycle engine or used to produce high-pressure steam. The present reference design is capable of producing about 2400 kW(t) (about 600 kW(e) net) for 15 full-power years. Technical and safety features are described along with recent progress in component hardware development programs and market assessment work.

Nuclear driven water decomposition plant for hydrogen production

NASA Technical Reports Server (NTRS)

Parker, G. H.; Brecher, L. E.; Farbman, G. H.

1976-01-01

The conceptual design of a hydrogen production plant using a very-high-temperature nuclear reactor (VHTR) to energize a hybrid electrolytic-thermochemical system for water decomposition has been prepared. A graphite-moderated helium-cooled VHTR is used to produce 1850 F gas for electric power generation and 1600 F process heat for the water-decomposition process which uses sulfur compounds and promises performance superior to normal water electrolysis or other published thermochemical processes. The combined cycle operates at an overall thermal efficiency in excess of 45%, and the overall economics of hydrogen production by this plant have been evaluated predicated on a consistent set of economic ground rules. The conceptual design and evaluation efforts have indicated that development of this type of nuclear-driven water-decomposition plant will permit large-scale economic generation of hydrogen in the 1990s.

Enhanced microwave absorption properties of epoxy composites containing graphite nanosheets@Fe3O4 decorated comb-like MnO2 nanoparticles

NASA Astrophysics Data System (ADS)

Su, Xiaogang; Wang, Jun; Zhang, Bin; Chen, Wei; Wu, Qilei; Dai, Wei; Zou, Yi

2018-05-01

Recently, owing to the radiation and interference from electromagnetic wave (EMW), the requirements of EMW absorbing materials have been increasing. Herein, a novel absorber composed of graphite nanosheets@Fe3O4 composites decorated comb-like MnO2 (GNFM) has been successfully synthesized via a facile two steps, characterized using x-ray diffraction (XRD), field emission scanning electron microscopy (FESEM), x-ray photoelectron spectroscopy (XPS), Raman spectroscopy, vibrating sample magnetometry (VSM) and vector network analyzer (VNA). The ternary composites with enhanced microwave absorption performance are due to the complementary effects of electroconductive material (graphite nanosheets), dielectric materials (MnO2) and magnetic material (Fe3O4 nanospheres). Hence, the maximum reflection loss of GNFM/epoxy composites is up to ‑31.7 dB at 5.85 GHz with absorbing thickness of 4.5 mm, and the efficient frequency bandwidth below ‑10 dB can reach up to 4.47 GHz (11.87–16.34 GHz) at matching thickness of 2 mm. The results demonstrate that GNFM could be regarded as a novel type of microwave absorbing material.

Observation of room-temperature high-energy resonant excitonic effects in graphene

NASA Astrophysics Data System (ADS)

Santoso, I.; Gogoi, P. K.; Su, H. B.; Huang, H.; Lu, Y.; Qi, D.; Chen, W.; Majidi, M. A.; Feng, Y. P.; Wee, A. T. S.; Loh, K. P.; Venkatesan, T.; Saichu, R. P.; Goos, A.; Kotlov, A.; Rübhausen, M.; Rusydi, A.

2011-08-01

Using a combination of ultraviolet-vacuum ultraviolet reflectivity and spectroscopic ellipsometry, we observe a resonant exciton at an unusually high energy of 6.3 eV in epitaxial graphene. Surprisingly, the resonant exciton occurs at room temperature and for a very large number of graphene layers N≈75, thus suggesting a poor screening in graphene. The optical conductivity (σ1) of a resonant exciton scales linearly with the number of graphene layers (up to at least 8 layers), implying the quantum character of electrons in graphene. Furthermore, a prominent excitation at 5.4 eV, which is a mixture of interband transitions from π to π* at the M point and a π plasmonic excitation, is observed. In contrast, for graphite the resonant exciton is not observable but strong interband transitions are seen instead. Supported by theoretical calculations, for N⩽ 28 the σ1 is dominated by the resonant exciton, while for N> 28 it is a mixture between exitonic and interband transitions. The latter is characteristic for graphite, indicating a crossover in the electronic structure. Our study shows that important elementary excitations in graphene occur at high binding energies and elucidate the differences in the way electrons interact in graphene and graphite.

NUCLEAR FISSION CHAIN REACTING SYSTEM

DOEpatents

Anderson, H.L.; Brown, H.S.

1961-06-27

The patent describes a reactor consisting of a plurality of tubes passing through a body of heavy water or graphite, a heat exchanger, means for flowing UF/sub 6/ through the tubes and the heat exchangar, and means for bleeding off some of the UF/sub 6/ and separating plutonium therefrom. A specific suggestion contained is that the amount of the UF/sub 6/ outside the reaction unit be a multiple of that within it.

A New {sup 14}C-AMS Facility at UFF- Niteroi, Brazil

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gomes, P. R. S.; Macario, K. D.; Anjos, R. M.

2010-08-04

We report a new Accelerator Mass Spectrometry facility at the Physics Institute of Fluminense Federal University in Brazil, the Nuclear Chronology Laboratory - LACRON. The sample preparation laboratory is ready to perform chemical treatment through graphitization and the acquisition of a Single Stage Accelerator Mass Spectrometry System is in progress. LACRON will be the first independent laboratory to perform the {sup 14}C-AMS technique not only in Brazil but in Latin America.

Advanced Thermally Stable Coal-Based Jet Fuels

DTIC Science & Technology

2008-01-01

the refined chemical oil used previously is in full agreement with recomm fuendations provided by RAND. Rxploratory work was begun to evaluate the...possible production of nuclear graphite from the by-products of co-coking. This would allow use of distillation residua rather than decant oil as feed...isotropic, that also means that the coker feed could be a distillation residuum (resid) and is not required to be a decant oil . Thus we envision the

Disclinations in Carbon-Carbon Composites.

DTIC Science & Technology

1983-09-01

8i-C-0641 U LASIFIED F/6G ii/4 N I uuuuullu ..D un n ." =25 1321. MICROCOP EOUINTSLHR NATONL = BUR A FSADRS16- UNCLASSI FI ED SECURITY CLASIrICA’sJM...Applications nuclear carbon carbon fiber intercalation compounds biocarbons and potential uses - Fundamentals physics chemistry technology The technical...Graphite intercalation compounds : old and new University of Munich problems in the chemist’s view West Germany L. S. Singer Carbon fibers from mesophase

A New 14C-AMS Facility at UFF- Niteroi, Brazil

NASA Astrophysics Data System (ADS)

Gomes, P. R. S.; Macario, K. D.; Anjos, R. M.; Linares, R.; Carvalho, C.; Queiroz, E.

2010-08-01

We report a new Accelerator Mass Spectrometry facility at the Physics Institute of Fluminense Federal University in Brazil, the Nuclear Chronology Laboratory—LACRON. The sample preparation laboratory is ready to perform chemical treatment through graphitization and the acquisition of a Single Stage Accelerator Mass Spectrometry System is in progress. LACRON will be the first independent laboratory to perform the 14C-AMS technique not only in Brazil but in Latin America.

Comparative Inhalation Screen of Titanium Dioxide and Graphite Dusts

DTIC Science & Technology

1988-11-01

refton.TTNX32pimn is recmmndfrtemnacueorfltiegmbds *OCALLAS:iumNO 302 pigmen id plcto notclo cat lse cotrledb usNAIE oNFLA LS:Aopnnftegasbth TITANOX 3020 pigment...58) described a similar glInds of affected wild rats contained infectious virus by disease of young rats, but inflammation was restricted to the...caused the disease. and Hunt detected acidophilic intra- studied, but virus and/or antibody have been detected in wild nuclear inclusion bodies in

Soft X-Ray Absorption Spectroscopy of High-Abrasion-Furnace Carbon Black

DOE Office of Scientific and Technical Information (OSTI.GOV)

Muramatsu, Yasuji; Harada, Ryusuke; Gullikson, Eric M.

2007-02-02

The soft x-ray absorption spectra of high-abrasion-furnace carbon black were measured to obtain local-structure/chemical-states information of the primary particles and/or crystallites. The soft x-ray absorption spectral features of carbon black represent broader {pi}* and {sigma}* peak structures compared to highly oriented pyrolytic graphite (HOPG). The subtracted spectra between the carbon black and HOPG, (carbon black) - (HOPG), show double-peak structures on both sides of the {pi}* peak. The lower-energy peak, denoted as the 'pre-peak', in the subtracted spectra and the {pi}*/{sigma}* peak intensity ratio in the absorption spectra clearly depend on the specific surface area by nitrogen adsorption (NSA). Therefore,more » it is concluded that the pre-peak intensity and the {pi}*/{sigma}* ratio reflect the local graphitic structure of carbon black.« less

Real-space Wigner-Seitz Cells Imaging of Potassium on Graphite via Elastic Atomic Manipulation

PubMed Central

Yin, Feng; Koskinen, Pekka; Kulju, Sampo; Akola, Jaakko; Palmer, Richard E.

2015-01-01

Atomic manipulation in the scanning tunnelling microscopy, conventionally a tool to build nanostructures one atom at a time, is here employed to enable the atomic-scale imaging of a model low-dimensional system. Specifically, we use low-temperature STM to investigate an ultra thin film (4 atomic layers) of potassium created by epitaxial growth on a graphite substrate. The STM images display an unexpected honeycomb feature, which corresponds to a real-space visualization of the Wigner-Seitz cells of the close-packed surface K atoms. Density functional simulations indicate that this behaviour arises from the elastic, tip-induced vertical manipulation of potassium atoms during imaging, i.e. elastic atomic manipulation, and reflects the ultrasoft properties of the surface under strain. The method may be generally applicable to other soft e.g. molecular or biomolecular systems. PMID:25651973

Thickness-dependent photocatalytic performance of graphite oxide for degrading organic pollutants under visible light.

PubMed

Oh, Junghoon; Chang, Yun Hee; Kim, Yong-Hyun; Park, Sungjin

2016-04-28

Photocatalysts use sustainable solar light energy to trigger various catalytic reactions. Metal-free nanomaterials have been suggested as cost-effective and environmentally friendly photocatalysts. In this work, we propose thickness-controlled graphite oxide (GO) as a metal-free photocatalyst, which is produced by exfoliating thick GO particles via stirring and sonication. All GO samples exhibit photocatalytic activity for degrading an organic pollutant, rhodamine B under visible light, and the thickest sample shows the best catalytic performance. UV-vis-NIR diffuse reflectance absorption spectra indicate that thicker GO samples absorb more vis-NIR light than thinner ones. Density-functional theory calculations show that GO has a much smaller band gap than that of single-layer graphene oxide, and thus suggest that the largely-reduced band gap is responsible for this trend of light absorption.

Fast Neutron Spectrum Potassium Worth for Space Power Reactor Design Validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Marshall, Margaret A.; Briggs, J. Blair

2015-03-01

A variety of critical experiments were constructed of enriched uranium metal (oralloy ) during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Ofmore » the hundreds of delayed critical experiments, one was performed that consisted of uranium metal annuli surrounding a potassium-filled, stainless steel can. The outer diameter of the annuli was approximately 13 inches (33.02 cm) with an inner diameter of 7 inches (17.78 cm). The diameter of the stainless steel can was 7 inches (17.78 cm). The critical height of the configurations was approximately 5.6 inches (14.224 cm). The uranium annulus consisted of multiple stacked rings, each with radial thicknesses of 1 inch (2.54 cm) and varying heights. A companion measurement was performed using empty stainless steel cans; the primary purpose of these experiments was to test the fast neutron cross sections of potassium as it was a candidate for coolant in some early space power reactor designs.The experimental measurements were performed on July 11, 1963, by J. T. Mihalczo and M. S. Wyatt (Ref. 1) with additional information in its corresponding logbook. Unreflected and unmoderated experiments with the same set of highly enriched uranium metal parts were performed at the Oak Ridge Critical Experiments Facility in the 1960s and are evaluated in the International Handbook for Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) with the identifier HEU MET FAST 051. Thin graphite reflected (2 inches or less) experiments also using the same set of highly enriched uranium metal parts are evaluated in HEU MET FAST 071. Polyethylene-reflected configurations are evaluated in HEU-MET-FAST-076. A stack of highly enriched metal discs with a thick beryllium top reflector is evaluated in HEU-MET-FAST-069, and two additional highly enriched uranium annuli with beryllium cores are evaluated in HEU-MET-FAST-059. Both detailed and simplified model specifications are provided in this evaluation. Both of these fast neutron spectra assemblies were determined to be acceptable benchmark experiments. The calculated eigenvalues for both the detailed and the simple benchmark models are within ~0.26 % of the benchmark values for Configuration 1 (calculations performed using MCNP6 with ENDF/B-VII.1 neutron cross section data), but under-calculate the benchmark values by ~7s because the uncertainty in the benchmark is very small: ~0.0004 (1s); for Configuration 2, the under-calculation is ~0.31 % and ~8s. Comparison of detailed and simple model calculations for the potassium worth measurement and potassium mass coefficient yield results approximately 70 – 80 % lower (~6s to 10s) than the benchmark values for the various nuclear data libraries utilized. Both the potassium worth and mass coefficient are also deemed to be acceptable benchmark experiment measurements.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhou, Xiangwen; Contescu, Cristian I.; Zhao, Xi

Mmore » atrix graphite (G) with incompletely graphitized binder used in high-temperature gas-cooled reactors (HTGRs) is commonly suspected to exhibit lower oxidation resistance in air. In order to reveal the oxidation performance, the oxidation behavior of newly developed A3-3 G at the temperature range from 500 to 950°C in air was studied and the effect of oxidation on the compressive strength of oxidized G specimens was characterized. Results show that temperature has a significant influence on the oxidation behavior of G. The transition temperature between Regimes I and II is ~700°C and the activation energy ( E a ) in Regime I is around 185 kJ/mol, a little lower than that of nuclear graphite, which indicates G is more vulnerable to oxidation. Oxidation at 550°C causes more damage to compressive strength of G than oxidation at 900°C. Comparing with the strength of pristine G specimens, the rate of compressive strength loss is 77.3% after oxidation at 550°C and only 12.5% for oxidation at 900°C. icrostructure images of SE and porosity measurement by ercury Porosimetry indicate that the significant compressive strength loss of G oxidized at 550°C may be attributed to both the uniform pore formation throughout the bulk and the preferential oxidation of the binder.« less

SIMPLIFIED SODIUM GRAPHITE REACTOR SYSTEM

DOEpatents

Dickinson, R.W.

1963-03-01

This patent relates to a nuclear power reactor comprising a reactor vessel, shielding means positioned at the top of said vessel, means sealing said reactor vessel to said shielding means, said vessel containing a quantity of sodium, a core tank, unclad graphite moderator disposed in said tank, means including a plurality of process tubes traversing said tank for isolating said graphite from said sodium, fuel elements positioned in said process tubes, said core tank being supported in spaced relation to the walls and bottom of said reactor vessel and below the level of said sodium, neutron shielding means positioned adjacent said core tank between said core tank and the walls of said vessel, said neutron shielding means defining an annuiar volume adjacent the inside wall of said reactor vessel, inlet plenum means below said core tank for providing a passage between said annular volume and said process tubes, heat exchanger means removably supported from the first-named shielding means and positioned in said annular volume, and means for circulating said sodium over said neutron shielding means down through said heat exchanger, across said inlet plenum and upward through said process tubes, said last-named means including electromagnetic pumps located outside said vessel and supported on said vessel wall between said heat exchanger means and said inlet plenum means. (AEC)

Nanocarbon: Defect Architectures and Properties

NASA Astrophysics Data System (ADS)

Vuong, Amanda

The allotropes of carbon make its solid phases amongst the most diverse of any element. It can occur naturally as graphite and diamond, which have very different properties that make them suitable for a wide range of technological and commercial purposes. Recent developments in synthetic carbon include Highly Oriented Pyrolytic Graphite (HOPG) and nano-carbons, such as fullerenes, nanotubes and graphene. The main industrial application of bulk graphite is as an electrode material in steel production, but in purified nuclear graphite form, it is also used as a moderator in Advanced Gas-cooled Reactors across the United Kingdom. Both graphene and graphite are damaged over time when subjected to bombardment by electrons, neutrons or ions, and these have a wide range of effects on their physical and electrical properties, depending on the radiation flux and temperature. This research focuses on intrinsic defects in graphene and dimensional change in nuclear graphite. The method used here is computational chemistry, which complements physical experiments. Techniques used comprise of density functional theory (DFT) and molecular dynamics (MD), which are discussed in chapter 2 and chapter 3, respectively. The succeeding chapters describe the results of simulations performed to model defects in graphene and graphite. Chapter 4 presents the results of ab initio DFT calculations performed to investigate vacancy complexes that are formed in AA stacked bilayer graphene. In AB stacking, carbon atoms surrounding the lattice vacancies can form interlayer structures with sp2 bonding that are lower in energy compared to in-plane reconstructions. From the investigation of AA stacking, sp2 interlayer bonding of adjacent multivacancy defects in registry creates a type of stable sp2 bonded wormhole between the layers. Also, a new class of mezzanine structure characterised by sp3 interlayer bonding, resembling a prismatic vacancy loop has also been identified. The mezzanine, which is a V6 hexavacancy variant, where six sp3 carbon atoms sit midway between two carbon layers and bond to both, is substantially more stable than any other vacancy aggregate in AA stacked layers. Chapter 5 presents the results of ab initio DFT calculations performed to investigate the wormhole and mezzanine defect that were identified in chapter 4 and the ramp defect discovered by Trevethan et al.. DFT calculations were performed on these defects in twisted bilayer graphene. From the investigation of vacancy complexes in twisted bilayer graphene, it is found that vacancy complexes are unstable in the twisted region and are more favourable in formation energy when the stacking arrangement is close to AA or AB stacking. It has also been discovered that the ramp defect is more stable in the twisted bilayer graphene compared to the mezzanine defect. Chapter 6 presents the results of ab initio DFT calculations performed to investigate a form of extending defect, prismatic edge dislocation. Suarez-Martinez et al.'s research suggest the armchair core is disconnected from any other layer, whilst the zigzag core is connected. In the investigation here, the curvature of the mezzanine defect allows it to swing between the armchair, zigzag and Klein in the AA stacking. For the AB stacking configuration, the armchair and zigzag core are connected from any other layer. Chapter 7 present results of MD simulations using the adaptive intermolecular reactive empirical bond order (AIREBO) potential to investigate the dimensional change of graphite due to the formation of vacancies present in a single crystal. It has been identified that there is an expansion along the c-axis, whilst a contraction along the a- and b- axes due to the coalescence of vacancy forming in-plane and between the layers. The results here are in good agreement with experimental studies of low temperature irradiation. The final chapter gives conclusions to this work.

Processing and fabrication of mixed uranium/refractory metal carbide fuels with liquid-phase sintering

NASA Astrophysics Data System (ADS)

Knight, Travis W.; Anghaie, Samim

2002-11-01

Optimization of powder processing techniques were sought for the fabrication of single-phase, solid-solution mixed uranium/refractory metal carbide nuclear fuels - namely (U, Zr, Nb)C. These advanced, ultra-high temperature nuclear fuels have great potential for improved performance over graphite matrix, dispersed fuels tested in the Rover/NERVA program of the 1960s and early 1970s. Hypostoichiometric fuel samples with carbon-to-metal ratios of 0.98, uranium metal mole fractions of 5% and 10%, and porosities less than 5% were fabricated. These qualities should provide for the longest life and highest performance capability for these fuels. Study and optimization of processing methods were necessary to provide the quality assurance of samples for meaningful testing and assessment of performance for nuclear thermal propulsion applications. The processing parameters and benefits of enhanced sintering by uranium carbide liquid-phase sintering were established for the rapid and effective consolidation and formation of a solid-solution mixed carbide nuclear fuel.

A Pulse Coupled Neural Network Segmentation Algorithm for Reflectance Confocal Images of Epithelial Tissue

PubMed Central

Malik, Bilal H.; Jabbour, Joey M.; Maitland, Kristen C.

2015-01-01

Automatic segmentation of nuclei in reflectance confocal microscopy images is critical for visualization and rapid quantification of nuclear-to-cytoplasmic ratio, a useful indicator of epithelial precancer. Reflectance confocal microscopy can provide three-dimensional imaging of epithelial tissue in vivo with sub-cellular resolution. Changes in nuclear density or nuclear-to-cytoplasmic ratio as a function of depth obtained from confocal images can be used to determine the presence or stage of epithelial cancers. However, low nuclear to background contrast, low resolution at greater imaging depths, and significant variation in reflectance signal of nuclei complicate segmentation required for quantification of nuclear-to-cytoplasmic ratio. Here, we present an automated segmentation method to segment nuclei in reflectance confocal images using a pulse coupled neural network algorithm, specifically a spiking cortical model, and an artificial neural network classifier. The segmentation algorithm was applied to an image model of nuclei with varying nuclear to background contrast. Greater than 90% of simulated nuclei were detected for contrast of 2.0 or greater. Confocal images of porcine and human oral mucosa were used to evaluate application to epithelial tissue. Segmentation accuracy was assessed using manual segmentation of nuclei as the gold standard. PMID:25816131

Revised Point of Departure Design Options for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Fittje, James E.; Borowski, Stanley K.; Schnitzler, Bruce

2015-01-01

In an effort to further refine potential point of departure nuclear thermal rocket engine designs, four proposed engine designs representing two thrust classes and utilizing two different fuel matrix types are designed and analyzed from both a neutronics and thermodynamic cycle perspective. Two of these nuclear rocket engine designs employ a tungsten and uranium dioxide cermet (ceramic-metal) fuel with a prismatic geometry based on the ANL-200 and the GE-710, while the other two designs utilize uranium-zirconium-carbide in a graphite composite fuel and a prismatic fuel element geometry developed during the Rover/NERVA Programs. Two engines are analyzed for each fuel type, a small criticality limited design and a 111 kN (25 klbf) thrust class engine design, which has been the focus of numerous manned mission studies, including NASA's Design Reference Architecture 5.0. slightly higher T/W ratios, but they required substantially more 235U.

Nuclear design analysis of square-lattice honeycomb space nuclear rocket engine

NASA Astrophysics Data System (ADS)

Widargo, Reza; Anghaie, Samim

1999-01-01

The square-lattice honeycomb reactor is designed based on a cylindrical core that is determined to have critical diameter and length of 0.50 m and 0.50 c, respectively. A 0.10-cm thick radial graphite reflector, in addition to a 0.20-m thick axial graphite reflector are used to reduce neutron leakage from the reactor. The core is fueled with solid solution of 93% enriched (U, Zr, Nb)C, which is one of several ternary uranium carbides that are considered for this concept. The fuel is to be fabricated as 2 mm grooved (U, Zr, Nb)C wafers. The fuel wafers are used to form square-lattice honeycomb fuel assemblies, 0.10 m in length with 30% cross-sectional flow area. Five fuel assemblies are stacked up axially to form the reactor core. Based on the 30% void fraction, the width of the square flow channel is about 1.3 mm. The hydrogen propellant is passed through these flow channels and removes the heat from the reactor core. To perform nuclear design analysis, a series of neutron transport and diffusion codes are used. The preliminary results are obtained using a simple four-group cross-section model. To optimize the nuclear design, the fuel densities are varied for each assembly. Tantalum, hafnium and tungsten are considered and used as a replacement for niobium in fuel material to provide water submersion sub-criticality for the reactor. Axial and radial neutron flux and power density distributions are calculated for the core. Results of the neutronic analysis indicate that the core has a relatively fast spectrum. From the results of the thermal hydraulic analyses, eight axial temperature zones are chosen for the calculation of group average cross-sections. An iterative process is conducted to couple the neutronic calculations with the thermal hydraulics calculations. Results of the nuclear design analysis indicate that a compact core can be designed based on ternary uranium carbide square-lattice honeycomb fuel. This design provides a relatively high thrust to weight ratio.

The WPI reactor-readying for the next generation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bobek, L.M.

1993-01-01

Built in 1959, the 10-kW open-pool nuclear training reactor at Worcester Polytechnic Institute (WPI) was one of the first such facilities in the nation located on a university campus. Since then, the reactor and its related facilities have been used to train two generations of nuclear engineers and scientists for the nuclear industry. With the use of nuclear technology playing an increasing role in many segments of the economy, WPI with its nuclear reactor facility is committed to continuing its mission of training future nuclear engineers and scientists. The WPI reactor includes a 6-in. beam port, graphite thermal column, andmore » in-core sample facility. The reactor, housed in an open 8000-gal tank of water, is designed so that the core is readily accessible. Both the control console and the peripheral counting equipment used for student projects and laboratory exercises are located in the reactor room. This arrangement provides convenience and flexibility in using the reactor for foil activations in neutron flux measurements, diffusion measurements, radioactive decay measurements, and the neutron activation of samples for analysis. In 1988, the reactor was successfully converted to low-enriched uranium fuel.« less

Can Silicon-Smelting Contribute to the Low O/Si Ratio on the Surface of Mercury?

NASA Technical Reports Server (NTRS)

McCubbin, F. M.; Vander Kaaden, K. E.; Hogancamp, J.; Archer, P. D., Jr.; Boyce, J. W.

2018-01-01

The MErcury Surface, Space ENvironment, GEochemistry, and Ranging (MESSENGER) spacecraft collected data that provided important insights into the structure, chemical makeup, and compositional diversity of Mercury. Among the many discoveries about Mercury made by MESSENGER, several surprising compositional characteristics of the surface were observed. These discoveries include elevated sulfur abundances (up to 4 wt.%), elevated abundances of graphitic carbon (0-4.1 wt.% across the surface with an additional 1-3 wt.% graphite above the global average in low reflectance materials), low iron abundances (less than 2 wt.%), and low oxygen abundances (O/Si weight ratio of 1.20+/-0.1). These exotic characteristics likely have important implications for the thermochemical evolution of Mercury and point to a planet that formed under highly reducing conditions. In the present study, we focus specifically on the low O/Si ratio of Mercury, which is anomalous compared to all other planetary materials. A recent study that considered the geochemical implications of the low O/Si ratio reported that 12-20% of the surface materials on Mercury are composed of Si-rich, Si-Fe alloys. They further postulated that the origin of the metal is best explained by a combination of space weathering and graphite-induced smelting that was facilitated by interaction of graphite with boninitic and komatiitic parental liquids. The goal of the present study is to assess the plausibility of smelting on Mercury through experiments run at the conditions that McCubbin et al. indicated would be favorable for Si-smelting.

Effect of Graphite Nanoplate Morphology on the Dispersion and Physical Properties of Polycarbonate Based Composites.

PubMed

Müller, Michael Thomas; Hilarius, Konrad; Liebscher, Marco; Lellinger, Dirk; Alig, Ingo; Pötschke, Petra

2017-05-18

The influence of the morphology of industrial graphite nanoplate (GNP) materials on their dispersion in polycarbonate (PC) is studied. Three GNP morphology types were identified, namely lamellar, fragmented or compact structure. The dispersion evolution of all GNP types in PC is similar with varying melt temperature, screw speed, or mixing time during melt mixing. Increased shear stress reduces the size of GNP primary structures, whereby the GNP aspect ratio decreases. A significant GNP exfoliation to individual or few graphene layers could not be achieved under the selected melt mixing conditions. The resulting GNP macrodispersion depends on the individual GNP morphology, particle sizes and bulk density and is clearly reflected in the composite's electrical, thermal, mechanical, and gas barrier properties. Based on a comparison with carbon nanotubes (CNT) and carbon black (CB), CNT are recommended in regard to electrical conductivity, whereas, for thermal conductive or gas barrier application, GNP is preferred.

Electron cyclotron resonance plasma reactor for production of carbon stripper foil

NASA Astrophysics Data System (ADS)

Faith Romero, Camille; Kanamori, Keita; Kinsho, Michikazu; Yoshimoto, Masahiro; Wada, Motoi

2018-01-01

A graphite antenna for the production of carbon-containing hydrogen plasmas is being developed to prepare impurity-free charge exchange foils for high-energy synchrotrons. Microwave power at 2.45 GHz frequency drives a coaxial structure antenna with a 12-mm-diameter central graphite cylinder and a tapered surrounding cylinder serving as the ground electrode. The antenna was placed in a linear magnetic field to investigate how it performs under an electron cyclotron resonance (ECR) condition. A clear resonance phenomenon was observed in plasma luminosity, microwave power absorption, and microwave power reflection when the induction current used to produce a linear magnetic field was changed. The antenna realized the best microwave coupling to the plasma with the ECR zone formed 5 mm from the end of the center electrode. The antenna realized stable operation for more than 5 h with 100 W input microwave power and with operating hydrogen pressure from 0.5 to 50 Pa.

Experimental and computational correlation of fracture parameters KIc, JIc, and GIc for unimodular and bimodular graphite components

NASA Astrophysics Data System (ADS)

Bhushan, Awani; Panda, S. K.

2018-05-01

The influence of bimodularity (different stress ∼ strain behaviour in tension and compression) on fracture behaviour of graphite specimens has been studied with fracture toughness (KIc), critical J-integral (JIc) and critical strain energy release rate (GIc) as the characterizing parameter. Bimodularity index (ratio of tensile Young's modulus to compression Young's modulus) of graphite specimens has been obtained from the normalized test data of tensile and compression experimentation. Single edge notch bend (SENB) testing of pre-cracked specimens from the same lot have been carried out as per ASTM standard D7779-11 to determine the peak load and critical fracture parameters KIc, GIc and JIc using digital image correlation technology of crack opening displacements. Weibull weakest link theory has been used to evaluate the mean peak load, Weibull modulus and goodness of fit employing two parameter least square method (LIN2), biased (MLE2-B) and unbiased (MLE2-U) maximum likelihood estimator. The stress dependent elasticity problem of three-dimensional crack progression behaviour for the bimodular graphite components has been solved as an iterative finite element procedure. The crack characterizing parameters critical stress intensity factor and critical strain energy release rate have been estimated with the help of Weibull distribution plot between peak loads versus cumulative probability of failure. Experimental and Computational fracture parameters have been compared qualitatively to describe the significance of bimodularity. The bimodular influence on fracture behaviour of SENB graphite has been reflected on the experimental evaluation of GIc values only, which has been found to be different from the calculated JIc values. Numerical evaluation of bimodular 3D J-integral value is found to be close to the GIc value whereas the unimodular 3D J-value is nearer to the JIc value. The significant difference between the unimodular JIc and bimodular GIc indicates that GIc should be considered as the standard fracture parameter for bimodular brittle specimens.

Dynamic friction and wear of a solid film lubricant during radiation exposure in a nuclear reactor

NASA Technical Reports Server (NTRS)

Jacobson, T. P.

1972-01-01

The effect of nuclear reactor radiation on the performance of a solid film lubricant was studied. The film consisted of molybdenum disulfide and graphite in a sodium silicate binder. Radiation levels of fast neutrons (E or = 1 MeV) were fluxed up to 3.5 times 10 to the 12th power n/sq cm-sec (intensity) and fluences up to 2 times 10 to the 18th power n/sq cm (total exposure). Coating wear lives were much shorter and friction coefficients higher in a high flux region of the reactor than in a low flux region. The amount of total exposure did not affect lubrication behavior as severely as the radiation intensity during sliding.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Larche, Michael R.; Prowant, Matthew S.; Bruillard, Paul J.

This study compares different approaches for imaging the internal architecture of graphite/epoxy composites using backscattered ultrasound. Two cases are studied. In the first, near-surface defects in a thin graphite/epoxy plates are imaged. The same backscattered waveforms were used to produce peak-to-peak, logarithm of signal energy, as well as entropy images of different types. All of the entropy images exhibit better border delineation and defect contrast than the either peak-to-peak or logarithm of signal energy. The best results are obtained using the joint entropy of the backscattered waveforms with a reference function. Two different references are examined. The first is amore » reflection of the insonifying pulse from a stainless steel reflector. The second is an approximate optimum obtained from an iterative parametric search. The joint entropy images produced using this reference exhibit three times the contrast obtained in previous studies. These plates were later destructively analyzed to determine size and location of near-surface defects and the results found to agree with the defect location and shape as indicated by the entropy images. In the second study, images of long carbon graphite fibers (50% by weight) in polypropylene thermoplastic are obtained as a first step toward ultrasonic determination of the distributions of fiber position and orientation.« less

Astrophysics with Presolar Stardust

NASA Astrophysics Data System (ADS)

Clayton, Donald D.; Nittler, Larry R.

2004-09-01

Meteorites and interplanetary dust particles contain presolar stardust grains: solid samples of stars that can be studied in the laboratory. The stellar origin of the grains is indicated by enormous isotopic ratio variations compared with Solar System materials, explainable only by nuclear reactions occurring in stars. Known presolar phases include diamond, SiC, graphite, Si3N4, Al2O3, MgAl2O4, CaAl12O19, TiO2, Mg(Cr,Al)2O4, and most recently, silicates. Subgrains of refractory carbides (e.g., TiC), and Fe-Ni metal have also been observed within individual presolar graphite grains. We review the astrophysical implications of these grains for the sciences of nucleosynthesis, stellar evolution, grain condensation, and the chemical and dynamic evolution of the Galaxy. Unique scientific information derives primarily from the high precision (in some cases <1%) of the measured isotopic ratios of large numbers of elements in single stardust grains. Stardust science is just now reaching maturity and will play an increasingly important role in nucleosynthesis applications.

Dual-ion-beam deposition of carbon films with diamond-like properties

NASA Technical Reports Server (NTRS)

Mirtich, M. J.; Swec, D. M.; Angus, J. C.

1985-01-01

A single and dual ion beam system was used to generate amorphous carbon films with diamond like properties. A methane/argon mixture at a molar ratio of 0.28 was ionized in the low pressure discharge chamber of a 30-cm-diameter ion source. A second ion source, 8 cm in diameter was used to direct a beam of 600 eV Argon ions on the substrates (fused silica or silicon) while the deposition from the 30-cm ion source was taking place. Nuclear reaction and combustion analysis indicate H/C ratios for the films to be 1.00. This high value of H/C, it is felt, allowed the films to have good transmittance. The films were impervious to reagents which dissolve graphitic and polymeric carbon structures. Although the measured density of the films was approximately 1.8 gm/cu cm, a value lower than diamond, the films exhibited other properties that were relatively close to diamond. These films were compared with diamond like films generated by sputtering a graphite target.

Dual ion beam deposition of carbon films with diamondlike properties

NASA Technical Reports Server (NTRS)

Mirtich, M. J.; Swec, D. M.; Angus, J. C.

1984-01-01

A single and dual ion beam system was used to generate amorphous carbon films with diamond like properties. A methane/argon mixture at a molar ratio of 0.28 was ionized in the low pressure discharge chamber of a 30-cm-diameter ion source. A second ion source, 8 cm in diameter was used to direct a beam of 600 eV Argon ions on the substrates (fused silica or silicon) while the deposition from the 30-cm ion source was taking place. Nuclear reaction and combustion analysis indicate H/C ratios for the films to be 1.00. This high value of H/C, it is felt, allowed the films to have good transmittance. The films were impervious to reagents which dissolve graphitic and polymeric carbon structures. Although the measured density of the films was approximately 1.8 gm/cu cm, a value lower than diamond, the films exhibited other properties that were relatively close to diamond. These films were compared with diamondlike films generated by sputtering a graphite target.

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

NASA Astrophysics Data System (ADS)

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

2018-03-01

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

Purification Procedures for Single-Wall Carbon Nanotubes

NASA Technical Reports Server (NTRS)

Gorelik, Olga P.; Nikolaev, Pavel; Arepalli, Sivaram

2001-01-01

This report summarizes the comparison of a variety of procedures used to purify carbon nanotubes. Carbon nanotube material is produced by the arc process and laser oven process. Most of the procedures are tested using laser-grown, single-wall nanotube (SWNT) material. The material is characterized at each step of the purification procedures by using different techniques including scanning electron microscopy (SEM), energy-dispersive X-ray spectroscopy (EDS), transmission electron microscopy (TEM), Raman, X-ray diffractometry (XRD), thermogravimetric analysis (TGA), nuclear magnetic resonance (NMR), and high-performance liquid chromatography (HPLC). The identified impurities are amorphous and graphitic carbon, catalyst particle aggregates, fullerenes, and hydrocarbons. Solvent extraction and low-temperature annealing are used to reduce the amount of volatile hydrocarbons and dissolve fullerenes. Metal catalysts and amorphous as well as graphitic carbon are oxidized by reflux in acids including HCl, HNO3 and HF and other oxidizers such as H2O2. High-temperature annealing in vacuum and in inert atmosphere helps to improve the quality of SWNTs by increasing crystallinity and reducing intercalation.

Magnesium fluoride reduction-vessel liners. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Latham-Brown, C.E.

1986-03-26

The work described in this report details a program that demonstrated a method by which magnesium fluoride, the by-product of the reduction reaction of uranium tetrafluoride to uranium metal could be used to replace the present graphite used to line the reduction vessel. Utilization of magnesium fluoride (MgF2) as a reduction-vessel liner has the potential to decrease carbon contamination and thereby reduce DU derby rejects due to chemistry. Additionally, there would be the elimination of the cost of the graphite crucible liner and the associated disposal costs by replacement with the by-product of the reduction reaction, which is magnesium fluoride.more » The process would ultimately result in reduced manufacturing costs for derby metal and higher yield of finished penetrators. This was to be accomplished in such a manner as to produce uranium metal derbies which would be accommodated into the present Nuclear Metals-Carolina Metals penetrator production process with minimal changes in equipment and procedures.« less

Effect of Vinylene Carbonate and Fluoroethylene Carbonate on SEI Formation on Graphitic Anodes in Li-Ion Batteries

DOE PAGES

Nie, Mengyun; Demeaux, Julien; Young, Benjamin T.; ...

2015-07-23

Binder free (BF) graphite electrodes were utilized to investigate the effect of electrolyte additives fluoroethylene carbonate (FEC) and vinylene carbonate (VC) on the structure of the solid electrolyte interface (SEI). The structure of the SEI has been investigated via ex-situ surface analysis including X-ray Photoelectron spectroscopy (XPS), Hard XPS (HAXPES), Infrared spectroscopy (IR) and transmission electron microscopy (TEM). The components of the SEI have been further investigated via nuclear magnetic resonance (NMR) spectroscopy of D2O extractions. The SEI generated on the BF-graphite anode with a standard electrolyte (1.2 M LiPF6 in ethylene carbonate (EC) / ethyl methyl carbonate (EMC), 3/7more » (v/v)) is composed primarily of lithium alkyl carbonates (LAC) and LiF. Incorporation of VC (3% wt) results in the generation of a thinner SEI composed of Li2CO3, poly(VC), LAC, and LiF. Incorporation of VC inhibits the generation of LAC and LiF. Incorporation of FEC (3% wt) also results in the generation of a thinner SEI composed of Li2CO3, poly(FEC), LAC, and LiF. The concentration of poly(FEC) is lower than the concentration of poly(VC) and the generation of LAC is inhibited in the presence of FEC. The SEI appears to be a homogeneous film for all electrolytes investigated.« less

Failure Predictions for VHTR Core Components using a Probabilistic Contiuum Damage Mechanics Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fok, Alex

2013-10-30

The proposed work addresses the key research need for the development of constitutive models and overall failure models for graphite and high temperature structural materials, with the long-term goal being to maximize the design life of the Next Generation Nuclear Plant (NGNP). To this end, the capability of a Continuum Damage Mechanics (CDM) model, which has been used successfully for modeling fracture of virgin graphite, will be extended as a predictive and design tool for the core components of the very high- temperature reactor (VHTR). Specifically, irradiation and environmental effects pertinent to the VHTR will be incorporated into the modelmore » to allow fracture of graphite and ceramic components under in-reactor conditions to be modeled explicitly using the finite element method. The model uses a combined stress-based and fracture mechanics-based failure criterion, so it can simulate both the initiation and propagation of cracks. Modern imaging techniques, such as x-ray computed tomography and digital image correlation, will be used during material testing to help define the baseline material damage parameters. Monte Carlo analysis will be performed to address inherent variations in material properties, the aim being to reduce the arbitrariness and uncertainties associated with the current statistical approach. The results can potentially contribute to the current development of American Society of Mechanical Engineers (ASME) codes for the design and construction of VHTR core components.« less

AGC-4 Experiment Irradiation Monitoring Data Qualification Interim Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hull, Laurence Charles

2016-08-01

The Graphite Technology Development Program is running a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The fourth experiment, Advanced Graphite Creep 4 (AGC 4), began with Advanced Test Reactor (ATR) cycle 157D on May 30, 2015, and has been irradiated for two cycles. The capsule was removed from the reactor after ATR cycle 158A, which ended on January 2, 2016, due to interference with another experiment. Irradiation will resume when the interfering experiment is removed from the reactor. This report documents qualification of AGC 4 experiment irradiation monitoring data for use by themore » Advanced Reactor Technologies (ART) Technology Development Office (TDO) Program for research and development activities required to design and license the first HTR nuclear plant. Qualified data meet the requirements for use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements and provide no useable information. Trend data may not meet all requirements, but still provide some useable information. Use of Trend data requires assessment of how any deficiencies affect a particular use of the data. All thermocouples (TCs) have functioned throughout the AGC-4 experiment. All temperature data are Qualified for use by the ART TDO Program. Argon, helium, and total gas flow data were within expected ranges and are Qualified for use by the ART TDO Program. Discharge gas line moisture values were consistently low during cycle 157D. At the start of cycle 158A, gas moisture briefly spiked to over 600 ppmv and then declined throughout the cycle. Moisture values are within the measurement range of the instrument and are Qualified for use by the ART TDO Program. Graphite creep specimens were subjected to one of three loads, 393, 491, or 589 lbf. For a brief period during cycle 157D between 12:19 on June 2, 2015 and 08:23 on June 11, 2015 the load cells were wired incorrectly resulting in missing stack load data. Missing stack loads were estimated from measured ram pressures using regression equations developed from the existing data from cycle 157D. Estimated stack loads during this period are considered to be an accurate representation of actual load applied to the stacks. These loads deviate slightly from the planned loads. This deviation does not prevent the data from being Qualified for use, but must be taken into account when analyzing the effect of load on creep. Stack displacement increased consistently throughout the first two cycles with total displacement ranging from 0.4 to 0.8 in. During ATR outages, a set of pneumatic rams raised the stacks of graphite creep specimens to ensure the specimens were not stuck within the test train. This stack raising was performed twice. All stacks were raised successfully each time. The load and displacement data are Qualified for use by the ART TDO Program.« less

Prepreg cure monitoring using diffuse reflectance-FTIR. [Fourier Transform Infrared Technique

NASA Technical Reports Server (NTRS)

Young, P. R.; Chang, A. C.

1984-01-01

An in situ diffuse reflectance-Fourier transform infrared technique was developed to determine infrared spectra of graphite fiber prepregs as they were being cured. A bismaleimide, an epoxy, and addition polyimide matrix resin prepregs were studied. An experimental polyimide adhesive was also examined. Samples were positioned on a small heater at the focal point of diffuse reflectance optics and programmed at 15 F/min while FTIR spectra were being scanned, averaged, and stored. An analysis of the resulting spectra provided basic insights into changes in matrix resin molecular structure which accompanied reactions such as imidization and crosslinking. An endo-exothermal isomerization involving reactive end-caps was confirmed for the addition polyimide prepregs. The results of this study contribute to a fundamental understanding of the processing of composites and adhesives. Such understanding will promote the development of more efficient cure cycles.

Measurement of thermal neutrons reflection coefficients for two-layer reflectors.

PubMed

Azimkhani, S; Zolfagharpour, F; Ziaie, F

2018-05-01

In this research, thermal neutrons albedo coefficients and relative number of excess counts have been measured experimentally for different thicknesses of two-layer reflectors by using 241 Am-Be neutron source (5.2Ci) and BF 3 detector. Our used reflectors consist of two-layer which are combinations of water, graphite, polyethylene, and lead materials. Experimental results reveal that thermal neutron reflection coefficients slightly increased by addition of the second layer. The maximum value of growth for thermal neutrons albedo is obtained for lead-polyethylene compound (0.72 ± 0.01). Also, there is suitable agreement between the experimental values and simulation results by using MCNPX code. Copyright © 2018 Elsevier Ltd. All rights reserved.

NERVA-Derived Nuclear Thermal Propulsion Dual Mode Operation

NASA Astrophysics Data System (ADS)

Zweig, Herbert R.; Hundal, Rolv

1994-07-01

Generation of electrical power using the nuclear heat source of a NERVA-derived nuclear thermal rocket engine is presented. A 111,200 N thrust engine defined in a study for NASA-LeRC in FY92 is the reference engine for a three-engine vehicle for which a 50 kWe capacity is required. Processes are described for energy extraction from the reactor and for converting the energy to electricity. The tie tubes which support the reactor fuel elements are the source of thermal energy. The study focuses on process systems using Stirling cycle energy conversion operating at 980 K and an alternate potassium-Rankine system operating at 1,140 K. Considerations are given of the effect of the power production on turbopump operation, ZrH moderator dissociation, creep strain in the tie tubes, hydrogen permeation through the containment materials, requirements for a backup battery system, and the effects of potential design changes on reactor size and criticality. Nuclear considerations include changing tie tube materials to TZM, changing the moderator to low vapor-pressure yttrium hydride, and changing the fuel form from graphite matrix to a carbon-carbide composite.

Method for producing dustless graphite spheres from waste graphite fines

DOEpatents

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.

Examples of Use of SINBAD Database for Nuclear Data and Code Validation

NASA Astrophysics Data System (ADS)

Kodeli, Ivan; Žerovnik, Gašper; Milocco, Alberto

2017-09-01

The SINBAD database currently contains compilations and evaluations of over 100 shielding benchmark experiments. The SINBAD database is widely used for code and data validation. Materials covered include: Air, N. O, H2O, Al, Be, Cu, graphite, concrete, Fe, stainless steel, Pb, Li, Ni, Nb, SiC, Na, W, V and mixtures thereof. Over 40 organisations from 14 countries and 2 international organisations have contributed data and work in support of SINBAD. Examples of the use of the database in the scope of different international projects, such as the Working Party on Evaluation Cooperation of the OECD and the European Fusion Programme demonstrate the merit and possible usage of the database for the validation of modern nuclear data evaluations and new computer codes.

Nursing education and the nuclear age

DOE Office of Scientific and Technical Information (OSTI.GOV)

McKay, S.

As reflected in the nursing literature, nurses have only recently begun discussing professional responsibilities for avoidance of nuclear war. The literature of the 1950s and 1960s focused on issues of civil defense. The 1970s were mostly silent, but with the onset of the 1980s a few articles identified the need for the nursing profession to recognize the importance of nuclear war prevention. The responsibility of nursing education for including content about nuclear issues has not been discussed in the professional literature. The author surveyed baccalaureate programs of nursing education to determine whether this lack of discussion was reflected in nursingmore » curricula. Responses indicated that the literature does not adequately reflect the level of activity and interest occurring within nursing education about nuclear issues. Nevertheless, because there is so little discussion in the professional literature, an implicit message is sent that nuclear issues are not of importance and that nurses should not openly address them.24 references.« less

FennoFlakes: a project for identifying flake graphite ores in the Fennoscandian shield and utilizing graphite in different applications

NASA Astrophysics Data System (ADS)

Palosaari, Jenny; Eklund, O.; Raunio, S.; Lindfors, T.; Latonen, R.-M.; Peltonen, J.; Smått, J.-H.; Kauppila, J.; Lund, S.; Sjöberg-Eerola, P.; Blomqvist, R.; Marmo, J.

2016-04-01

Natural graphite is a strategic mineral, since the European Commission stated (Report on critical raw materials for the EU (2014)) that graphite is one of the 20 most critical materials for the European Union. The EU consumed 13% of all flake graphite in the world but produced only 3%, which stresses the demand of the material. Flake graphite, which is a flaky version of graphite, forms under high metamorphic conditions. Flake graphite is important in different applications like batteries, carbon brushes, heat sinks etc. Graphene (a single layer of graphite) can be produced from graphite and is commonly used in many nanotechnological applications, e.g. in electronics and sensors. The steps to obtain pure graphene from graphite ore include fragmentation, flotation and exfoliation, which can be cumbersome and resulting in damaging the graphene layers. We have started a project named FennoFlakes, which is a co-operation between geologists and chemists to fill the whole value chain from graphite to graphene: 1. Exploration of graphite ores (geological and geophysical methods). 2. Petrological and geochemical analyses on the ores. 3. Development of fragmentation methods for graphite ores. 4. Chemical exfoliation of the enriched flake graphite to separate flake graphite into single and multilayer graphene. 5. Test the quality of the produced material in several high-end applications with totally environmental friendly and disposable material combinations. Preliminary results show that flake graphite in high metamorphic areas has better qualities compared to synthetic graphite produced in laboratories.

A mathematical model for the release of noble gas and Cs from porous nuclear fuel based on VEGA 1&2 experiments

NASA Astrophysics Data System (ADS)

Simones, M. P.; Reinig, M. L.; Loyalka, S. K.

2014-05-01

Release of fission products from nuclear fuel in accidents is an issue of major concern in nuclear reactor safety, and there is considerable room for development of improved models, supported by experiments, as one needs to understand and elucidate role of various phenomena and parameters. The VEGA (Verification Experiments of radionuclides Gas/Aerosol release) program on several irradiated nuclear fuels investigated the release rates of radionuclides and results demonstrated that the release rates of radionuclides from all nuclear fuels tested decreased with increasing external gas pressure surrounding the fuel. Hidaka et al. (2004-2011) accounted for this pressure effect by developing a 2-stage diffusion model describing the transport of radionuclides in porous nuclear fuel. We have extended this 2-stage diffusion model to account for mutual binary gas diffusion in the open pores as well as to introduce the appropriate parameters to cover the slip flow regime (0.01 ⩽ Kn ⩽ 0.1). While we have directed our numerical efforts toward the simulation of the VEGA experiments and assessments of differences from the results of Hidaka et al., the model and the techniques reported here are of larger interest as these would aid in modeling of diffusion in general (e.g. in graphite and other nuclear materials of interest).

What the Erythrocytic Nuclear Alteration Frequencies Could Tell Us about Genotoxicity and Macrophage Iron Storage?

PubMed

Gomes, Juliana M M; Ribeiro, Heder J; Procópio, Marcela S; Alvarenga, Betânia M; Castro, Antônio C S; Dutra, Walderez O; da Silva, José B B; Corrêa Junior, José D

2015-01-01

Erythrocytic nuclear alterations have been considered as an indicative of organism's exposure to genotoxic agents. Due to their close relationship among their frequencies and DNA damages, they are considered excellent markers of exposure in eukaryotes. However, poor data has been found in literature concerning their genesis, differential occurrence and their life span. In this study, we use markers of cell viability; genotoxicity and cellular turn over in order to shed light to these events. Tilapia and their blood were exposed to cadmium in acute exposure and in vitro assays. They were analyzed using flow cytometry for oxidative stress and membrane disruption, optical microscopy for erythrocytic nuclear alteration, graphite furnace atomic absorption spectrometry for cadmium content in aquaria water, blood and cytochemical and analytical electron microscopy techniques for the hemocateretic aspects. The results showed a close relationship among the total nuclear alterations and cadmium content in the total blood and melanomacrophage centres area, mismatching reactive oxygen species and membrane damages. Moreover, nuclear alterations frequencies (vacuolated, condensed and blebbed) showed to be associated to cadmium exposure whereas others (lobed and bud) were associated to depuration period. Decrease on nuclear alterations frequencies was also associated with hemosiderin increase inside spleen and head kidney macrophages mainly during depurative processes. These data disclosure in temporal fashion the main processes that drive the nuclear alterations frequencies and their relationship with some cellular and systemic biomarkers.

What the Erythrocytic Nuclear Alteration Frequencies Could Tell Us about Genotoxicity and Macrophage Iron Storage?

PubMed Central

Gomes, Juliana M. M.; Ribeiro, Heder J.; Procópio, Marcela S.; Alvarenga, Betânia M.; Castro, Antônio C. S.; Dutra, Walderez O.; da Silva, José B. B.; Corrêa Junior, José D.

2015-01-01

Erythrocytic nuclear alterations have been considered as an indicative of organism’s exposure to genotoxic agents. Due to their close relationship among their frequencies and DNA damages, they are considered excellent markers of exposure in eukaryotes. However, poor data has been found in literature concerning their genesis, differential occurrence and their life span. In this study, we use markers of cell viability; genotoxicity and cellular turn over in order to shed light to these events. Tilapia and their blood were exposed to cadmium in acute exposure and in vitro assays. They were analyzed using flow cytometry for oxidative stress and membrane disruption, optical microscopy for erythrocytic nuclear alteration, graphite furnace atomic absorption spectrometry for cadmium content in aquaria water, blood and cytochemical and analytical electron microscopy techniques for the hemocateretic aspects. The results showed a close relationship among the total nuclear alterations and cadmium content in the total blood and melanomacrophage centres area, mismatching reactive oxygen species and membrane damages. Moreover, nuclear alterations frequencies (vacuolated, condensed and blebbed) showed to be associated to cadmium exposure whereas others (lobed and bud) were associated to depuration period. Decrease on nuclear alterations frequencies was also associated with hemosiderin increase inside spleen and head kidney macrophages mainly during depurative processes. These data disclosure in temporal fashion the main processes that drive the nuclear alterations frequencies and their relationship with some cellular and systemic biomarkers. PMID:26619141

ENGINEERING AND CONSTRUCTING THE HALLAM NUCLEAR POWER FACILITY REACTOR STRUCTURE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mahlmeister, J E; Haberer, W V; Casey, D F

1960-12-15

The Hallam Nuclear Power Facility reactor structure, including the cavity liner, is described, and the design philosophy and special design requirements which were developed during the preliminary and final engineering phases of the project are explained. The structure was designed for 600 deg F inlet and 1000 deg F outlet operating sodium temperatures and fabricated of austenitic and ferritic stainless steels. Support for the reactor core components and adequate containment for biological safeguards were readily provided even though quite conservative design philosophy was used. The calculated operating characteristics, including heat generation, temperature distributions and stress levels for full-power operation, aremore » summarized. Ship fabrication and field installation experiences are also briefly related. Results of this project have established that the sodium graphite reactor permits practical and economical fabrication and field erection procedures; considerably higher operating design temperatures are believed possible without radical design changes. Also, larger reactor structures can be similarly constructed for higher capacity (300 to 1000 Mwe) nuclear power plants. (auth)« less

Seal welded cast iron nuclear waste container

DOEpatents

Filippi, Arthur M.; Sprecace, Richard P.

1987-01-01

This invention identifies methods and articles designed to circumvent metallurgical problems associated with hermetically closing an all cast iron nuclear waste package by welding. It involves welding nickel-carbon alloy inserts which are bonded to the mating plug and main body components of the package. The welding inserts might be bonded in place during casting of the package components. When the waste package closure weld is made, the most severe thermal effects of the process are restricted to the nickel-carbon insert material which is far better able to accommodate them than is cast iron. Use of nickel-carbon weld inserts should eliminate any need for pre-weld and post-weld heat treatments which are a problem to apply to nuclear waste packages. Although the waste package closure weld approach described results in a dissimilar metal combination, the relative surface area of nickel-to-iron, their electrochemical relationship, and the presence of graphite in both materials will act to prevent any galvanic corrosion problem.

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

PubMed

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.

Carbon nanotube-based black coatings

NASA Astrophysics Data System (ADS)

Lehman, J.; Yung, C.; Tomlin, N.; Conklin, D.; Stephens, M.

2018-03-01

Coatings comprising carbon nanotubes are very black, that is, characterized by uniformly low reflectance over a broad range of wavelengths from the visible to far infrared. Arguably, there is no other material that is comparable. This is attributable to the intrinsic properties of graphitic material as well as the morphology (density, thickness, disorder, and tube size). We briefly describe a history of other coatings such as nickel phosphorous, gold black, and carbon-based paints and the comparable structural morphology that we associate with very black coatings. The need for black coatings is persistent for a variety of applications ranging from baffles and traps to blackbodies and thermal detectors. Applications for space-based instruments are of interest and we present a review of space qualification and the results of outgassing measurements. Questions of nanoparticle safety depend on the nanotube size and aspect ratio as well as the nature and route of exposure. We describe the growth of carbon nanotube forests along with the catalyst requirements and temperature limitations. We also describe coatings derived from carbon nanotubes and applied like paint. Building the measurement apparatus and determining the optical properties of something having negligible reflectance are challenging and we summarize the methods and means for such measurements. There exists information in the literature for effective media approximations to model the dielectric function of vertically aligned arrays. We summarize this along with the refractive index of graphite from the literature that is necessary for modeling the optical properties. In our experience, the scientific questions can be overshadowed by practical matters, so we provide an appendix of recipes for making as-grown and sprayed coatings along with an example of reflectance measurements.

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

NASA Astrophysics Data System (ADS)

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.

Presolar stardust in meteorites: recent advances and scientific frontiers

NASA Astrophysics Data System (ADS)

Nittler, Larry R.

2003-04-01

Grains of stardust that formed in stellar outflows prior to the formation of the solar system survive intact as trace constituents of primitive meteorites. The presolar origin of the grains is indicated by enormous isotopic ratio variations compared to solar system materials. Identified presolar phases include diamond, silicon carbide, graphite, silicon nitride, corundum, spinel, hibonite, titanium oxide, and, most recently, silicates. Sub-grains of refractory carbides (e.g. TiC), and Fe-Ni metal have also been observed within individual presolar graphite grains. Isotopic compositions indicate that the grains formed in red giants, asymptotic giant branch (AGB) stars, supernovae and novae; thus they provide unique insights into the evolution of and nucleosynthesis within these environments. Some of the isotopic variations also reflect the chemical evolution of the galaxy and can be used to constrain corresponding models. Presolar grain microstructures provide information about physical and chemical conditions of dust formation in stellar environments; recent studies have focused on graphite grains from supernovae as well as SiC and corundum from AGB stars. The survival of presolar grains in different classes of meteorites has important implications for early solar system evolution. Recent analytical developments, including resonance ionization mass spectrometry, high spatial resolution secondary ion mass spectrometry and site-selective ion milling, should help solve many outstanding problems but are likely to also introduce new surprises.

Producing graphite with desired properties

NASA Technical Reports Server (NTRS)

Dickinson, J. M.; Imprescia, R. J.; Reiswig, R. D.; Smith, M. C.

1971-01-01

Isotropic or anisotropic graphite is synthesized with precise control of particle size, distribution, and shape. The isotropic graphites are nearly perfectly isotropic, with thermal expansion coefficients two or three times those of ordinary graphites. The anisotropic graphites approach the anisotropy of pyrolytic graphite.

Brazing graphite to graphite

DOEpatents

Peterson, George 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 virtually graphite.

AGC-2 Specimen Post Irradiation Data Package Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Windes, William Enoch; Swank, W. David; Rohrbaugh, David T.

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

Method of Joining Graphite Fibers to a Substrate

NASA Technical Reports Server (NTRS)

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.

Carbonization in Titan Tholins: implication for low albedo on surfaces of Centaurs and trans-Neptunian objects

NASA Astrophysics Data System (ADS)

Giri, Chaitanya; McKay, Christopher P.; Goesmann, Fred; Schäfer, Nadine; Li, Xiang; Steininger, Harald; Brinckerhoff, William B.; Gautier, Thomas; Reitner, Joachim; Meierhenrich, Uwe J.

2016-07-01

Astronomical observations of Centaurs and trans-Neptunian objects (TNOs) yield two characteristic features - near-infrared (NIR) reflectance and low geometric albedo. The first feature apparently originates due to complex organic material on their surfaces, but the origin of the material contributing to low albedo is not well understood. Titan tholins synthesized to simulate aerosols in the atmosphere of Saturn's moon Titan have also been used for simulating the NIR reflectances of several Centaurs and TNOs. Here, we report novel detections of large polycyclic aromatic hydrocarbons, nanoscopic soot aggregates and cauliflower-like graphite within Titan tholins. We put forth a proof of concept stating the surfaces of Centaurs and TNOs may perhaps comprise of highly `carbonized' complex organic material, analogous to the tholins we investigated. Such material would apparently be capable of contributing to the NIR reflectances and to the low geometric albedos simultaneously.

Raman spectral characteristics of magmatic-contact metamorphic coals from Huainan Coalfield, China

NASA Astrophysics Data System (ADS)

Chen, Shancheng; Wu, Dun; Liu, Guijian; Sun, Ruoyu

2017-01-01

Normal burial metamorphism of coal superimposed by magmatic-contact metamorphism makes the characteristics of the Raman spectrum of coal changed. Nine coal samples were chosen at a coal transect perpendicular to the intrusive dike, at the No. 3 coal seam, Zhuji Coal Mine, Huainan Coalfield, China, with different distances from dike-coal boundary (DCB). Geochemical (proximate and ultimate) analysis and mean random vitrinite reflectance (R0, %) indicate that there is a significant relationship between the values of volatile matter and R0 in metamorphosed coals. Raman spectra show that the graphite band (G band) becomes the major band but the disordered band (D band) disappears progressively, with the increase of metamorphic temperature in coals, showing that the structural organization in high-rank contact-metamorphosed coals is close to that of well-crystallized graphite. Evident relationships are observed between the calculated Raman spectral parameters and the peak metamorphic temperature, suggesting some spectral parameters have the potentials to be used as geothermometers for contact-metamorphic coals.

Effect of Graphite Nanoplate Morphology on the Dispersion and Physical Properties of Polycarbonate Based Composites

PubMed Central

Müller, Michael Thomas; Hilarius, Konrad; Liebscher, Marco; Lellinger, Dirk; Alig, Ingo; Pötschke, Petra

2017-01-01

The influence of the morphology of industrial graphite nanoplate (GNP) materials on their dispersion in polycarbonate (PC) is studied. Three GNP morphology types were identified, namely lamellar, fragmented or compact structure. The dispersion evolution of all GNP types in PC is similar with varying melt temperature, screw speed, or mixing time during melt mixing. Increased shear stress reduces the size of GNP primary structures, whereby the GNP aspect ratio decreases. A significant GNP exfoliation to individual or few graphene layers could not be achieved under the selected melt mixing conditions. The resulting GNP macrodispersion depends on the individual GNP morphology, particle sizes and bulk density and is clearly reflected in the composite’s electrical, thermal, mechanical, and gas barrier properties. Based on a comparison with carbon nanotubes (CNT) and carbon black (CB), CNT are recommended in regard to electrical conductivity, whereas, for thermal conductive or gas barrier application, GNP is preferred. PMID:28772907

Fabrication of near-net shape graphite/magnesium composites for large mirrors

NASA Astrophysics Data System (ADS)

Wendt, Robert; Misra, Mohan

1990-10-01

Successful development of space-based surveillance and laser systems will require large precision mirrors which are dimensionally stable under thermal, static, and dynamic (i.e., structural vibrations and retargeting) loading conditions. Among the advanced composites under consideration for large space mirrors, graphite fiber reinforced magnesium (Gr/Mg) is an ideal candidate material that can be tailored to obtain an optimum combination of properties, including a high modulus of elasticity, zero coefficient of thermal expansion, low density, and high thermal conductivity. In addition, an innovative technique, combining conventional filament winding and vacuum casting has been developed to produce near-net shape Gr/Mg composites. This approach can significantly reduce the cost of fabricating large mirrors by decreasing required machining. However, since Gr/Mg cannot be polished to a reflective surface, plating is required. This paper will review research at Martin Marietta Astronautics Group on Gr/Mg mirror blank fabrication and measured mechanical and thermal properties. Also, copper plating and polishing methods, and optical surface characteristics will be presented.

Ray propagation path analysis of acousto-ultrasonic signals in composites

NASA Technical Reports Server (NTRS)

Kautz, Harold E.

1987-01-01

The most important result was the demonstration that acousto-ultrasonic (AU) energy introduced into a laminated graphite/resin propagates by two modes through the structure. The first mode, along the graphite fibers, is the faster. The second mode, through the resin matrix, besides being slower is also more strongly attenuated at the higher frequencies. This demonstration was accomplished by analyzing the time and frequency domain of the composite AU signal and comparing them to the same for a neat resin specimen of the same chemistry and geometry as the composite matrix. Analysis of the fine structure of AU spectra was accomplished by various geometrical strategies. It was shown that the multitude of narrow peaks associated with AU spectra are the effect of the many pulse arrivals in the signal. The shape and distribution of the peaks is mainly determined by the condition of nonnormal reflections of ray paths. A cepstrum analysis was employed which can be useful in detecting characteristic times. Analysis of propagation modes can be accomplished while ignoring the fine structure.

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

NASA Astrophysics Data System (ADS)

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

2013-02-01

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

Formation of dysprosium carbide on the graphite (0001) surface

DOE PAGES

Lii-Rosales, Ann; Zhou, Yinghui; Wallingford, Mark; ...

2017-07-12

When using scanning tunneling microscopy, we characterize a surface carbide that forms such that Dy is deposited on the basal plane of graphite. In order to form carbide islands on terraces, Dy is first deposited at 650–800 K, which forms large metallic islands. Upon annealing at 1000 K, these clusters convert to carbide. Deposition directly at 1000 K is ineffective because nucleation on terraces is inhibited. Reaction is signaled by the fact that each carbide cluster is partially or totally surrounded by an etch pit. The etch pit is one carbon layer deep for most carbide clusters. Carbide clusters aremore » also identifiable by striations on their surfaces. Based on mass balance, and assuming that only the surface layer of carbon is involved in the reaction, the carbide has stoichiometry D y 2 C . This is Dy-rich compared with the most common bulk carbide Dy C 2 , which may reflect limited surface carbon transport to the carbide.« less

Performance and Ageing Robustness of Graphite/NMC Pouch Prototypes Manufactured through Eco-Friendly Materials and Processes.

PubMed

Loeffler, Nicholas; Kim, Guk-T; Passerini, Stefano; Gutierrez, Cesar; Cendoya, Iosu; De Meatza, Iratxe; Alessandrini, Fabrizio; Appetecchi, Giovanni B

2017-09-22

Graphite/lithium nickel-manganese-cobalt oxide (NMC), stacked pouch cells with nominal capacity of 15-18 Ah were designed, developed, and manufactured for automotive applications in the frame of the European Project GREENLION. A natural, water-soluble material was used as the main electrode binder, thus allowing the employment of H 2 O as the only processing solvent. The electrode formulations were developed, optimized, and upscaled for cell manufacturing. Prolonged cycling and ageing tests revealed excellent capacity retention and robustness toward degradation phenomena. For instance, above 99 % of the initial capacity is retained upon 500 full charge/discharge cycles, corresponding to a fading of 0.004 % per cycle, and about 80 % of the initial capacity is delivered after 8 months ageing at 45 °C. The stacked soft-packaged cells have shown very reproducible characteristics and performance, reflecting the goodness of design and manufacturing. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Advanced materials for space nuclear power systems

NASA Technical Reports Server (NTRS)

Titran, Robert H.; Grobstein, Toni L.; Ellis, David L.

1991-01-01

The overall philosophy of the research was to develop and characterize new high temperature power conversion and radiator materials and to provide spacecraft designers with material selection options and design information. Research on three candidate materials (carbide strengthened niobium alloy PWC-11 for fuel cladding, graphite fiber reinforced copper matrix composites for heat rejection fins, and tungsten fiber reinforced niobium matrix composites for fuel containment and structural supports considered for space power system applications is discussed. Each of these types of materials offers unique advantages for space power applications.

Thermally exfoliated graphite oxide

NASA Technical Reports Server (NTRS)

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.

Stress wave attenuation in thin structures by ultrasonic through-transmission

NASA Technical Reports Server (NTRS)

Lee, S. S.; Williams, J. H., Jr.

1980-01-01

The steady state amplitude of the output of an ultrasonic through transmission measurement is analyzed and the result is given in closed form. Provided that the product of the input and output transduction ratios; the specimen-transducer reflection coefficient; the specimen-transducer phase shift parameter; and the material phase velocity are known, this analysis gives a means for determining the through-thickness attenuation of an individual thin sample. Multiple stress wave reflections are taken into account and so signal echoes do not represent a difficulty. An example is presented for a graphite fiber epoxy composite (Hercules AS/3501-6). A direct method for continuous or intermittent monitoring of through thickness attenuation of plate structures which may be subject to service structural degradation is provided.

The action of macrosounds on graphite ore and derived products

NASA Technical Reports Server (NTRS)

Bradeteanu, C.; Dragan, O.

1974-01-01

A suspension of graphite ore, floated graphite, and the gangue left over from flotation were subjected to the action of macrosounds under determinant conditions. The following was found: (1) The graphite ore undergoes an efficient settling action. (2) The floated graphite is strongly crushed down to the dimensions of colloidal graphite. (3) The gangue left over from flotation can be further processed to recuperate graphite from its nuclei.

Graphene prepared by thermal reduction–exfoliation of graphite oxide: Effect of raw graphite particle size on the properties of graphite oxide and graphene

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dao, Trung Dung; Jeong, Han Mo, E-mail: hmjeong@mail.ulsan.ac.kr

Highlights: • Effect of raw graphite particle size on properties of GO and graphene is reported. • Size of raw graphite affects oxidation degree and chemical structure of GO. • Highly oxidized GO results in small-sized but well-exfoliated graphene. • GO properties affect reduction degree, structure, and conductivity of graphene. - Abstract: We report the effect of raw graphite size on the properties of graphite oxide and graphene prepared by thermal reduction–exfoliation of graphite oxide. Transmission electron microscope analysis shows that the lateral size of graphene becomes smaller when smaller size graphite is used. X-ray diffraction analysis confirms that graphitemore » with smaller size is more effectively oxidized, resulting in a more effective subsequent exfoliation of the obtained graphite oxide toward graphene. X-ray photoelectron spectroscopy demonstrates that reduction of the graphite oxide derived from smaller size graphite into graphene is more efficient. However, Raman analysis suggests that the average size of the in-plane sp{sup 2}-carbon domains on graphene is smaller when smaller size graphite is used. The enhanced reduction degree and the reduced size of sp{sup 2}-carbon domains contribute contradictively to the electrical conductivity of graphene when the particle size of raw graphite reduces.« less

Enhanced performance of graphite anode materials by AlF3 coating for lithium-ion batteries

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ding, Fei; Xu, Wu; Choi, Daiwon

2012-04-27

In order to form the stable surface film and to further enhance the long-term cycling stability of the graphite anodes of lithium-ion batteries, the surface of graphite powders has been modified by AlF3 coating through chemical precipitation method. The AlF3-coated graphite shows no evident changes in the bulk structure and a thin AlF3-coating layer of about 2 nm thick is found to uniformly cover the graphite particles with 2 wt% AlF3 content. However, it delivers a higher initial discharge capacity and largely improved rate performances compared to the pristine graphite. Remarkably, AlF3 coated graphite demonstrated a much better cycle life.more » After 300 cycles, AlF3 coated graphite and uncoated graphite show capacity retention of 92% and 81%, respectively. XPS measurement shows that a more conductive solid electrode interface (SEI) layer was formed on AlF3 coated graphite as compared to uncoated graphite. SEM monograph also reveals that the AlF3-coated graphite particles have a much more stable surface morphology after long-term cycling. Therefore, the improved electrochemical performance of AlF3 coated graphite can be attributed to a more stable and conductive SEI formed on coated graphite anode during cycling process.« less

Graphite

USGS Publications Warehouse

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 China’s graphite production is expected to increase, although rising labor costs and some mine production problems are developing. China is expected to continue to be the dominant exporter for the near future. Mexico and Canada export graphite mainly to the United States, which has not had domestic production of natural graphite since the 1950s. Most graphite deposits in the United States are too small, low-grade, or remote to be of commercial value in the near future, and the likelihood of discovering larger, higher-grade, or favorably located domestic deposits is unlikely. The United States is a major producer of synthetic graphite.

Strategic graphite, a survey

USGS Publications Warehouse

Cameron, Eugene N.; Weis, Paul L.

1960-01-01

Strategic graphite consists of certain grades of lump and flake graphite for which the United States is largely or entirely dependent on sources abroad. Lump graphite of high purity, necessary in the manufacture of carbon brushes, is imported from Ceylon, where it occurs in vein deposits. Flake graphite, obtained from deposits consisting of graphite disseminated in schists and other metamorphic rocks, is an essential ingredient of crucibles used in the nonferrous metal industries and in the manufacture of lubricants and packings. High-quality flake graphite for these uses has been obtained mostly from Madagascar since World War I. Some flake graphite of strategic grade has been produced, however, from deposits in Texas, Alabama, and Pennsylvania. The development of the carbon-bonded crucible, which does not require coarse flake, should lessen the competitive advantage of the Madagascar producers of crucible flake. Graphite of various grades has been produced intermittently in the United States since 1644. The principal domestic deposits of flake graphite are in Texas, Alabama, Pennsylvania, and New York. Reserves of flake graphite in these four States are very large, but production has been sporadic and on the whole unprofitable since World War I, owing principally to competition from producers in Madagascar. Deposits in Madagascar are large and relatively high in content of flake graphite. Production costs are low and the flake produced is of high quality. Coarseness of flake and uniformity of the graphite products marketed are cited as major advantages of Madagascar flake. In addition, the usability of Madagascar flake for various purposes has been thoroughly demonstrated, whereas the usability of domestic flake for strategic purposes is still in question. Domestic graphite deposits are of five kinds: deposits consisting of graphite disseminated in metamorphosed siliceous sediments, deposits consisting of graphite disseminated in marble, deposits formed by thermal or dynamothermal metamorphism of coal beds or other highly carbonaceous sediments, vein deposits, and contact metasomatic deposits in marble. Only the first kind comprises deposits sufficiently large and rich in flake graphite to be significant potential sources of strategic grades of graphite. Vein deposits in several localities are known, but none is known to contain substantial reserves of graphite of strategic quality.Large resources of flake graphite exist in central Texas, in northeastern Alabama, in eastern Pennsylvania, and in the eastern Adirondack Mountains of New York. Tonnages available, compared with the tonnages of flake graphite consumed annually in the United States, are very large. There have been indications that flake graphite from Texas, Alabama, and Pennsylvania can be used in clay-graphite crucibles as a substitute for Madagascar flake, and one producer has made progress in establishing markets for his flake products as ingredients of lubricants. The tonnages of various commercial grades of graphite recoverable from various domestic deposits, however, have not been established; hence, the adequacy of domestic resources of graphite in a time of emergency is not known.The only vein deposits from which significant quantities of lump graphite have been produced are those of the Crystal Graphite mine, Beaverhead County, Mont. The deposits are fracture fillings in Precambrian gneiss and pegmatite. Known reserves in the deposits are small. In Texas, numerous flake-graphite deposits occur in the Precambrian Packsaddle schist in Llano and Burnet Counties. Graphite disseminated in certain parts of this formation ranges from extremely fine to medium grained. The principal producer has been the mine of the Southwestern Graphite Co., west of the town of Burnet. Substantial reserves of medium-grained graphite are present in the deposit mined by the company. In northeastern Alabama, flake-graphite deposits occur in the Ashland mica schist in two belts that trend northeastward across Clay, Goosa, and Chilton Counties. The northeastern belt has been the most productive. About 40 mines have been operated at one time or another, but only a few have been active during or since World War I. The deposits consist of flake graphite disseminated in certain zones or "leads" consisting of quartz-mica-feldspar schists and mica quartzite. Most of past production has come from the weathered upper parts of the deposits, but unweathered rock has been mined at several localities. Reserves of weathered rock containing 3 to 5 percent graphite are very large, and reserves of unweathered rock are even greater. Flake graphite deposits in Chester County, Pa., have been worked intermittently since about 1890. The deposits consist of medium- to coarse-grained graphite disseminated in certain belts of the Pickering gneiss. The most promising deposit is one worked in the Benjamin Franklin and the Eynon Just mines. Reserves of weathered rock containing 1.5 percent graphite are of moderate size; reserves of unweathered rock are large. In the eastern Adirondack Mountains in New York there are two principal kinds of flake-graphite deposits: contact-metasomatic deposits and those consisting of flake graphite disseminated in quartz schist. The contact-metasomatic deposits are small, irregular, and very erratic in graphite content. The deposits in quartz schist are very large, persistent, and uniform in grade. There are large reserves of schist containing 3 to 5 percent graphite, but the graphite is relatively fine grained.

The early development of neutron diffraction: Science in the wings of the Manhattan Project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mason, Thom; Gawne, Timothy J; Nagler, Stephen E

2012-01-01

Although neutron diffraction was first observed using radioactive decay sources shortly after the discovery of the neutron, it was only with the availability of higher intensity neutron beams from the first nuclear reactors, constructed as part of the Manhattan project, that systematic investigation of Bragg scattering became possible. Remarkably, at a time when the war effort was singularly focused on the development of the atomic bomb, groups working at Oak Ridge and Chicago carried out key measurements and recognized the future utility of neutron diffraction quite independent of its contributions to the measurements of nuclear cross sections. Ernest O. Wollan,more » Lyle B. Borst, and Walter H. Zinn were all able to observe neutron diffraction in 1944 using the X-10 graphite reactor and the CP-3 heavy water reactor.« less

Monte Carlo simulation of the operational quantities at the realistic mixed neutron-photon radiation fields CANEL and SIGMA.

PubMed

Lacoste, V; Gressier, V

2007-01-01

The Institute for Radiological Protection and Nuclear Safety owns two facilities producing realistic mixed neutron-photon radiation fields, CANEL, an accelerator driven moderator modular device, and SIGMA, a graphite moderated americium-beryllium assembly. These fields are representative of some of those encountered at nuclear workplaces, and the corresponding facilities are designed and used for calibration of various instruments, such as survey meters, personal dosimeters or spectrometric devices. In the framework of the European project EVIDOS, irradiations of personal dosimeters were performed at CANEL and SIGMA. Monte Carlo calculations were performed to estimate the reference values of the personal dose equivalent at both facilities. The Hp(10) values were calculated for three different angular positions, 0 degrees, 45 degrees and 75 degrees, of an ICRU phantom located at the position of irradiation.

CHAIN REACTING SYSTEM

DOEpatents

Fermi, E.; Leverett, M.C.

1958-06-01

A nuclear reactor of the gas-cooled, graphitemoderated type is described. In this design, graphite blocks are arranged in a substantially cylindrical lattice having vertically orienied coolant channels in which uranium fuel elements having through passages are disposed. The active lattice is contained within a hollow body. such as a steel shell, which, in turn, is surrounded by water and concrete shields. Helium is used as the primary coolant and is circulated under pressure through the coolant channels and fuel elements. The helium is then conveyed to heat exchangers, where its heat is used to produce steam for driving a prime mover, thence to filtering means where radioactive impurities are removed. From the filtering means the helium passes to a compressor and an after cooler and is ultimately returned to the reactor for recirculation. Control and safety rods are provided to stabilize or stop the reaction. A space is provided between the graphite lattice and the internal walls of the shell to allow for thermal expansion of the lattice during operation. This space is filled with a resilient packing, such as asbestos, to prevent the passage of helium.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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

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

DOE PAGES

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

2017-12-14

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

CMB-13 research on carbon and graphite

NASA Technical Reports Server (NTRS)

Smith, M. C.

1972-01-01

Preliminary results of the research on carbon and graphite accomplished during this report period are presented. Included are: particle characteristics of Santa Maria fillers, compositions and density data for hot-molded Santa Maria graphites, properties of hot-molded Santa Maria graphites, and properties of hot-molded anisotropic graphites. Ablation-resistant graphites are also discussed.

Stable dispersions of polymer-coated graphitic nanoplatelets

NASA Technical Reports Server (NTRS)

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

2011-01-01

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

Structural disorder of graphite and implications for graphite thermometry

NASA Astrophysics Data System (ADS)

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.

A SPACESHIP WITH NUCLEAR PROPULSION

DOE Office of Scientific and Technical Information (OSTI.GOV)

Polorny, J.

1962-01-01

ABS>A proposed space vehicle with nuclear propulsion for a round-trip Martian mission is described. It would be powered by a 270-Mw graphite- moderated, U-fueled nuclear reactor with a core 1 m high by 1 m in diameter, and use gas as propellant. The gas would be heated to the maximum temperature in the reactor and additionally accelerated by an electromagnetic field. To this end, small quantities of K would be injected into the gas stream to increase its electric conductivity. The required electrical energy would be produced by liquid-Na-cooled thermionic converters. The vehicle would weigh 115000 kg, including 43000 kgmore » of H propellant with tankage, and 7000 kg of sustenance material for one year. Chemical rockets would launch the vehicle with a crew of three men into an earth orbit where nuclear propulsion would take over. Upon reactor start-up, three heat exchangers (minimum dimensions 30 x 18 m) would be fanned out. A shielded well with a diameter of 2.5 m would protect the crew from radiation during reactor operation, passage through the earth radiation belts, and at periods of solar flares. (OTS)« less

International strategic minerals inventory summary report; natural graphite

USGS Publications Warehouse

Krauss, U.H.; Schmidt, H.W.; Taylor, H.A.; Sutphin, D.M.

1989-01-01

Natural graphite is a crystalline mineral of pure carbon which normally occurs in the form of platelet-shaped crystals. It has important properties, such as chemical inertness, low thermal expansion, and lubricity, that make it almost irreplaceable for certain uses such as refractories and steelmaking. Graphite ore types are crystalline (flake and lump} or 'amorphous' (cryptocrystalline}. Refractory applications use the largest total amount of natural graphite, while the most important use of crystalline graphite is in crucibles for handling molten metals. All graphite deposits being mined today are found in the following metamorphic environments: (1) contact metamorphosed coal generally is a source of amorphous graphite; (2)disseminated crystalline flake graphite comes from syngenetic metasediments; and (3) crystalline lump graphite is found in epigenetic veins in high-grade metamorphic regions. Graphite may also occur as a trace mineral in ultrabasic rocks and pegmatites, but these are economically insignificant. The world's identified economically exploitable resources of crystalline graphite in major deposits are estimated to be about 9.7 million metric tons of concentrate. In-place resources of amorphous graphite are about 11.5 million metric tons. Of these, less than 2 percent of the crystalline ore and less than 1 percent of the amorphous ore are in western industrial countries. World mining production of natural graphite rose from 347,000 metric tons in 1973 to 659,000 metric tons in 1986, while the proportion produced by central economy countries increased from about 50 percent for the period from 1973 to 1978 to more than 64 percent in 1979 to 1986. It is estimated that crystalline flake graphite accounts for at least 180,000 metric tons of total annual world mining production of natural graphite, and amorphous graphite makes up the rest.

Transformation of graphite by tectonic and hydrothermal processes in an active plate boundary fault zone, Alpine Fault, New Zealand

NASA Astrophysics Data System (ADS)

Kirilova, Matina; Toy, Virginia; Timms, Nicholas; Halfpenny, Angela; Menzies, Catriona; Craw, Dave; Rooney, Jeremy; Giorgetti, Carolina

2017-04-01

Graphite is a material with one of the lowest frictional strengths, with coefficient of friction of 0.1 and thus in natural fault zones it may act as a natural solid lubricant. Graphitization, or the transformation of organic matter (carbonaceous material, or CM) into crystalline graphite, is induced by compositional and structural changes during diagenesis and metamorphism. The supposed irreversible nature of this process has allowed the degree of graphite crystallinity to be calibrated as an indicator of the peak temperatures reached during progressive metamorphism. We examine processes of graphite emplacement and deformation in the Alpine Fault Zone, New Zealand's active continental tectonic plate boundary. Raman spectrometry indicates that graphite in the distal, amphibolite-facies Alpine Schist, which experienced peak metamorphic temperatures up to 640 ◦C, is highly crystalline and occurs mainly along grain boundaries within quartzo-feldspathic domains. The subsequent mylonitisation in the Alpine Fault Zone resulted in progressive reworking of CM under lower temperature conditions (500◦C-600◦C) in a structurally controlled environment, resulting in spatial clustering in lower-strain protomylonites, and further foliation-alignment in higher-strain mylonites. Subsequent brittle deformation of the mylonitised schists resulted in cataclasites that contain over three-fold increase in the abundance of graphite than mylonites. Furthermore, cataclasites contain graphite with two different habits: highly-crystalline, foliated forms that are inherited mylonitic graphite; and lower-crystallinity, less mature patches of finer-grained graphite. The observed graphite enrichment and the occurrence of poorly-organised graphite in the Alpine Fault cataclasites could result from: i) hydrothermal precipitation from carbon-supersaturated fluids; and/or ii) mechanical degradation by structural disordering of mylonitic graphite combined with strain-induced graphite localisation. The lack of published systematic studies of mechanical modification of the structure of graphite inhibits further conclusion to be drawn. Thus, we performed laboratory deformation experiments during which we sheared highly crystalline graphite powder at room temperature, normal stresses of 5 MPa and 25 MPa and sliding velocities of 1 µm/s, 10 µm/s and 100 µm/s. The degree of graphite crystallinity, both in the starting and resulting materials, was analysed by Raman microspectroscopy. Our results demonstrate consistent decrease of graphite crystallinity with increasing shear strain. We conclude that: i) graphite 'thermometers' are unreliable in brittely deformed rocks; ii) a shear strain calibration of graphite 'thermometers' is needed; iii) fault creep is very likely responsible for the observed structural and textural characteristics of graphite in the Alpine Fault cataclasites. Finally, to investigate the possibility of hydrothermal origin for at least some of the graphite in the Alpine Fault cataclasites we will also present synchrotron FTIR and carbon isotope analysis of the Alpine fault rocks.

Space Nuclear Facility test capability at the Baikal-1 and IGR sites Semipalatinsk-21, Kazakhstan

NASA Astrophysics Data System (ADS)

Hill, T. J.; Stanley, M. L.; Martinell, J. S.

1993-01-01

The International Space Technology Assessment Program was established 1/19/92 to take advantage of the availability of Russian space technology and hardware. DOE had two delegations visit CIS and assess its space nuclear power and propulsion technologies. The visit coincided with the Conference on Nuclear Power Engineering in Space Nuclear Rocket Engines at Semipalatinsk-21 (Kurchatov, Kazakhstan) on Sept. 22-25, 1992. Reactor facilities assessed in Semipalatinski-21 included the IVG-1 reactor (a nuclear furnace, which has been modified and now called IVG-1M), the RA reactor, and the Impulse Graphite Reactor (IGR), the CIS version of TREAT. Although the reactor facilities are being maintained satisfactorily, the support infrastructure appears to be degrading. The group assessment is based on two half-day tours of the Baikals-1 test facility and a brief (2 hr) tour of IGR; because of limited time and the large size of the tour group, it was impossible to obtain answers to all prepared questions. Potential benefit is that CIS fuels and facilities may permit USA to conduct a lower priced space nuclear propulsion program while achieving higher performance capability faster, and immediate access to test facilities that cannot be available in this country for 5 years. Information needs to be obtained about available data acquisition capability, accuracy, frequency response, and number of channels. Potential areas of interest with broad application in the U.S. nuclear industry are listed.

EXPLORATORY DEVELOPMENT OF GRAPHITE MATERIALS.

DTIC Science & Technology

COMPOSITE MATERIALS), (* GRAPHITE , (*FIBERS, GRAPHITE ), (*LAMINATED PLASTICS, GRAPHITE ), MOLDINGS, EXTRUSION, VACUUM, EPOXY RESINS, FILAMENTS, STRESSES, TENSILE PROPERTIES, OXIDATION, PHYSICAL PROPERTIES.

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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 canmore » 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.« less

40 CFR 436.380 - Applicability; description of the graphite subcategory.

Code of Federal Regulations, 2012 CFR

2012-07-01

... graphite subcategory. 436.380 Section 436.380 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY... Graphite Subcategory § 436.380 Applicability; description of the graphite subcategory. The provisions of this subpart are applicable to the mining and processing of naturally occurring graphite. ...

Method for producing thin graphite flakes with large aspect ratios

DOEpatents

Bunnell, L. Roy

1993-01-01

A method for making graphite flakes of high aspect ratio by the steps of providing a strong concentrated acid and heating the graphite in the presence of the acid for a time and at a temperature effective to intercalate the acid in the graphite; heating the intercalated graphite at a rate and to a temperature effective to exfoliate the graphite in discrete layers; subjecting the graphite layers to ultrasonic energy, mechanical shear forces, or freezing in an amount effective to separate the layes into discrete flakes.

Physical, electrochemical, and thermal properties of granulated natural graphite as anodes for Li-ion batteries.

PubMed

Jo, Yong Nam; Park, Min-Sik; Kim, Jae-Hun; Kim, Young-Jun

2013-05-01

Two different types of granulated graphites were synthesized by blending and kneading of natural graphite with pitch followed by sintering methods. The electrochemical performances of granulated graphites were investigated as anode materials for use in Li-ion batteries. The blending type granulated graphite possesses a large amount of cavities and voids, while the kneading type granulated graphite has a relatively compact microstructure, which is responsible for a high tap density. Both granulated graphites show improved the initial coulombic efficiencies as a result of decrease of surface area by the granulations. In particular, the kneading type granulated graphite exhibits an excellent rate-capability without significant capacity loss. In addition, the thermal stabilities of both granulated graphites were also improved, which could be attributed to the decrease of active surface area due to pitch coating.

Systems and methods for forming defects on graphitic materials and curing radiation-damaged graphitic materials

DOEpatents

Ryu, Sunmin; Brus, Louis E.; Steigerwald, Michael L.; Liu, Haitao

2012-09-25

Systems and methods are disclosed herein for forming defects on graphitic materials. The methods for forming defects include applying a radiation reactive material on a graphitic material, irradiating the applied radiation reactive material to produce a reactive species, and permitting the reactive species to react with the graphitic material to form defects. Additionally, disclosed are methods for removing defects on graphitic materials.

Andreev reflection without Fermi surface alignment in high- T c van der Waals heterostructures

DOE PAGES

Zareapour, Parisa; Hayat, Alex; Zhao, Shu Yang F.; ...

2017-04-05

We address the controversy over the proximity effect between topological materials and high-T c superconductors. Junctions are produced between Bi 2Sr 2CaCu 2Omore » $${}_{8+\\delta }$$ and materials with different Fermi surfaces (Bi 2Te 3 and graphite). Both cases reveal tunneling spectra that are consistent with Andreev reflection. This is confirmed by a magnetic field that shifts features via the Doppler effect. This is modeled with a single parameter that accounts for tunneling into a screening supercurrent. Thus the tunneling involves Cooper pairs crossing the heterostructure, showing that the Fermi surface mismatch does not hinder the ability to form transparent interfaces, which is accounted for by the extended Brillouin zone and different lattice symmetries.« less

Refurbishing of carbon contaminated pre-mirror of reflectivity beam line at Indus-1

NASA Astrophysics Data System (ADS)

Yadav, P. K.; Kumar, M.; Gupta, R. K.; Sinha, M.; Patel, H. S.; Modi, M. H.

2018-04-01

In recent days optics contamination and its refurbishing is a serious issue for synchrotron radiation beam line community. Here we refurbished a carbon contaminated mirror by Ar and O2 gas mixed (1:1) radio frequency plasma. For structural analysis pre and post characterization of the mirror was done by Soft X-ray reflectivity (SXRR), Raman Spectroscopy (RS) and Atomic force microscopy (AFM). Before refurbishing mirror, a low density graphitic carbon layer of thickness 400 Å with surface roughness about 55 Å and Au surface roughness 14Å was estimated by SXRR. After one hour RF plasma exposure it is observed by SXRR and Raman spectroscopy that carbon layer is completely removed. The AFM and SXRR results show that roughness of Au surface not increase after plasma exposure.

The Role of Carbon in Core Formation Under Highly Reducing Conditions With Implications for the Planet Mercury

NASA Technical Reports Server (NTRS)

Vander Kaaden, Kathleen E..; McCubbin, Francis M.; Ross, D. Kent; Draper, David S.

2017-01-01

Results from the MErcury Surface, Space ENvironment, GEochemistry and Ranging (MESSENGER) spacecraft have shown elevated abundances of carbon on the surface of Mercury. Furthermore, the X-Ray Spectrometer on board MESSENGER measured elevated abundances of sulfur and low abundances of iron, suggesting the planet's oxygen fugacity (fO2) is several log10 units below the Iron-Wüstite (IW) buffer. Similar to the role of other volatiles (e.g. sulfur) on highly reducing planetary bodies, carbon is expected to behave differently than it would under higher fO2. As discussed by Nittler et al. and Hauck et al., under such highly reducing conditions, the majority of the iron partitions into the core. On Mercury, this resulted in a relatively large core and a thin mantle. Using a composition similar to the largest volcanic field on the planet (the northern volcanic plains), Vander Kaaden and McCubbin conducted sink-float experiments to determine the density of melts and minerals on Mercury. They showed that graphite would be the only buoyant mineral in a mercurian magma ocean. Therefore, Vander Kaaden and McCubbin proposed a possible primary flotation crust on the planet composed of graphite. Concurrently, Peplowski et al. used GRS data from MESSENGER to show an average northern hemisphere abundance of C on the planet of 1.4 +/- 0.9 wt%. However, as this result was only at the one-sigma detection limit, possible carbon abundances at the three-sigma detection limit for Mercury range from 0 to 4.1 wt% carbon. Additionally, Murchie et al. investigated the possible darkening agent on Mercury and concluded that coarse-grained graphite could darken high reflectance plains to the low reflectance material. To further test the possibility of elevated abundances of carbon in Mercury's crust, Peplowski et al. used the low-altitude MESSENGER data to show that carbon is the only material consistent with both the visible to near-infrared spectra and the neutron measurements of low reflectance material on Mercury, confirming that C is the primary darkening agent on Mercury. Confirmation of carbon on the planet prompts many questions regarding the role of carbon during the differentiation and evolution of Mercury. Given the elevated abundances of both S and C on Mercury's surface, it begs the question, what is the core composition of the planet? This study seeks to understand the impact of C as a light element on potential core compositions on Mercury.

Levamisole: a Common Adulterant in Cocaine Street Samples Hindering Electrochemical Detection of Cocaine.

PubMed

de Jong, Mats; Florea, Anca; Vries, Anne-Mare de; van Nuijs, Alexander L N; Covaci, Adrian; Van Durme, Filip; Martins, José C; Samyn, Nele; De Wael, Karolien

2018-04-17

The present work investigates the electrochemical determination of cocaine in the presence of levamisole, one of the most common adulterants found in cocaine street samples. Levamisole misleads cocaine color tests, giving a blue color (positive test) even in the absence of cocaine. Moreover, the electrochemical detection of cocaine is also affected by the presence of levamisole, with a suppression of the oxidation signal of cocaine. When levamisole is present in the sample in ratios higher than 1:1, the cocaine signal is no longer detected, thus leading to false negative results. Mass spectrometry and nuclear magnetic resonance were used to investigate if the signal suppression is due to the formation of a complex between cocaine and levamisole in bulk solution. Strategies to eliminate this suppressing effect are further suggested in this manuscript. In a first approach, the increase of the pH of the sample solution from pH 7 to pH 12 allowed the voltammetric determination of cocaine in the presence of levamisole in a concentration range from 10 to 5000 μM at nonmodified graphite disposable electrodes with a detection limit of 5 μM. In a second approach, the graphite electrode was cathodically pretreated, resulting in the presence of oxidation peaks of both cocaine and levamisole, with a detection limit for cocaine of 3 μM over the linear range of concentrations from 10 to 2500 μM. Both these strategies have been successfully applied for the simultaneous detection of cocaine and levamisole in three street samples on unmodified graphite disposable electrodes.

Relaxation processes and glass transition of confined polymer melts: A molecular dynamics simulation of 1,4-polybutadiene between graphite walls.

PubMed

Solar, M; Binder, K; Paul, W

2017-05-28

Molecular dynamics simulations of a chemically realistic model for 1,4-polybutadiene in a thin film geometry confined by two graphite walls are presented. Previous work on melts in the bulk has shown that the model faithfully reproduces static and dynamic properties of the real material over a wide temperature range. The present work studies how these properties change due to nano-confinement. The focus is on orientational correlations observable in nuclear magnetic resonance experiments and on the local intermediate incoherent neutron scattering function, F s (q z , z, t), for distances z from the graphite walls in the range of a few nanometers. Temperatures from about 2T g down to about 1.15T g , where T g is the glass transition temperature in the bulk, are studied. It is shown that weakly attractive forces between the wall atoms and the monomers suffice to effectively bind a polymer coil that is near the wall. For a wide regime of temperatures, the Arrhenius-like adsorption/desorption kinetics of the monomers is the slowest process, while very close to T g the Vogel-Fulcher-Tammann-like α-relaxation takes over. The α-process is modified only for z≤1.2 nm due to the density changes near the walls, less than expected from studies of coarse-grained (bead-spring-type) models. The weakness of the surface effects on the glass transition in this case is attributed to the interplay of density changes near the wall with the torsional potential. A brief discussion of pertinent experiments is given.

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

NASA Astrophysics Data System (ADS)

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.

Preparation, quantitative surface analysis, intercalation characteristics and industrial implications of low temperature expandable graphite

NASA Astrophysics Data System (ADS)

Peng, Tiefeng; Liu, Bin; Gao, Xuechao; Luo, Liqun; Sun, Hongjuan

2018-06-01

Expandable graphite is widely used as a new functional carbon material, especially as fire-retardant; however, its practical application is limited due to the high expansion temperature. In this work, preparation process of low temperature and highly expandable graphite was studied, using natural flake graphite as raw material and KMnO4/HClO4/NH4NO3 as oxidative intercalations. The structure, morphology, functional groups and thermal properties were characterized during expanding process by Fourier transform infrared spectroscopy (FTIR), Raman spectra, thermo-gravimetry differential scanning calorimetry (TG-DSC), X-ray diffraction (XRD), and scanning electron microscope (SEM). The analysis showed that by oxidation intercalation, some oxygen-containing groups were grafted on the edge and within the graphite layer. The intercalation reagent entered the graphite layer to increase the interlayer spacing. After expansion, the original flaky expandable graphite was completely transformed into worm-like expanded graphite. The order of graphite intercalation compounds (GICs) was proposed and determined to be 3 for the prepared expandable graphite, based on quantitative XRD peak analysis. Meanwhile, the detailed intercalation mechanisms were also proposed. The comprehensive investigation paved a benchmark for the industrial application of such sulfur-free expanded graphite.

The impact of LDEF results on the space application of metal matrix composites

NASA Technical Reports Server (NTRS)

Steckel, Gary L.; Le, Tuyen D.

1993-01-01

Over 200 graphite/aluminum and graphite/magnesium composites were flown on the leading and trailing edges of LDEF on the Advanced Composites Experiment. The performance of these composites was evaluated by performing scanning electron microscopy and x-ray photoelectron spectroscopy of exposed surfaces, optical microscopy of cross sections, and on-orbit and postflight thermal expansion measurements. Graphite/aluminum and graphite/magnesium were found to be superior to graphite/polymer matrix composites in that they are inherently resistant to atomic oxygen and are less susceptible to thermal cycling induced microcracking. The surface foils on graphite/aluminum and graphite/magnesium protect the graphite fibers from atomic oxygen and from impact damage from small micrometeoroid or space debris particles. However, the surface foils were found to be susceptible to thermal fatigue cracking arising from contamination embrittlement, surface oxidation, or stress risers. Thus, the experiment reinforced requirements for carefully protecting these composites from prelaunch oxidation or corrosion, avoiding spacecraft contamination, and designing composite structures to minimize stress concentrations. On-orbit strain measurements demonstrated the importance of through-thickness thermal conductivity in composites to minimize thermal distortions arising from thermal gradients. Because of the high thermal conductivity of aluminum, thermal distortions were greatly reduced in the LDEF thermal environment for graphite/aluminum as compared to graphite/magnesium and graphite/polymer composites. The thermal expansion behavior of graphite/aluminum and graphite/magnesium was stabilized by on-orbit thermal cycling in the same manner as observed in laboratory tests.

Natural graphite demand and supply - Implications for electric vehicle battery requirements

USGS Publications Warehouse

Olson, Donald W.; Virta, Robert L.; Mahdavi, Mahbood; Sangine, Elizabeth S.; Fortier, Steven M.

2016-01-01

Electric vehicles have been promoted to reduce greenhouse gas emissions and lessen U.S. dependence on petroleum for transportation. Growth in U.S. sales of electric vehicles has been hindered by technical difficulties and the high cost of the lithium-ion batteries used to power many electric vehicles (more than 50% of the vehicle cost). Groundbreaking has begun for a lithium-ion battery factory in Nevada that, at capacity, could manufacture enough batteries to power 500,000 electric vehicles of various types and provide economies of scale to reduce the cost of batteries. Currently, primary synthetic graphite derived from petroleum coke is used in the anode of most lithium-ion batteries. An alternate may be the use of natural flake graphite, which would result in estimated graphite cost reductions of more than US$400 per vehicle at 2013 prices. Most natural flake graphite is sourced from China, the world's leading graphite producer. Sourcing natural flake graphite from deposits in North America could reduce raw material transportation costs and, given China's growing internal demand for flake graphite for its industries and ongoing environmental, labor, and mining issues, may ensure a more reliable and environmentally conscious supply of graphite. North America has flake graphite resources, and Canada is currently a producer, but most new mining projects in the United States require more than 10 yr to reach production, and demand could exceed supplies of flake graphite. Natural flake graphite may serve only to supplement synthetic graphite, at least for the short-term outlook.

Organic tissues, graphite, and hydrocarbons in host rocks of the Rum Jungle Uranium Field, northern Australia

USGS Publications Warehouse

Foster, C.B.; Robbins, E.I.; Bone, Y.

1990-01-01

The Rum Jungle Uranium field consists of at least six early Proterozoic deposits that have been mined either for uranium and/or the associated base and precious metals. Organic matter in the host rocks of the Whites Formation and Coomalie Dolomite is now predominantly graphite, consistent with the metamorphic history of these rocks. For nine samples, the mean total organic carbon content is high (3.9 wt%) and ranged from 0.33 to 10.44 wt%. Palynological extracts from the host rocks include black, filamentous, stellate (Eoastrion-like), and spherical morphotypes, which are typical of early Proterozoic microbiota. The colour, abundance, and shapes of these morphotypes reflect the thermal history, organic richness, and probable lacustrine biofacies of the host rocks. Routine analysis of rock thin sections and of palynological residues shows that mineral grains in some of the host rocks are coated with graphitized organic matter. The grain coating is presumed to result from ultimate thermal degradation of a petroleum phase that existed prior to metamorphism. Hydrocarbons are, however, still present in fluid inclusions within carbonates of the Coomalie Dolomite and lower Whites Formation. The fluid inclusions fluoresce dull orange in blue-light excitation and their hydrocarbon content is confirmed by gas chromatography of whole-rock extracts. Preliminary analysis of the oil suggests that it is migrated, and because it has escaped graphitization through metamorphism it is probably not of early Proterozoic age. The presence of live oil is consistent with fluid inclusion data that suggest subsequent, low-temperature brine migration through the rocks. The present observations support earlier suggestions that organic matter in the host formations trapped uranium to form protore. Subsequent fluid migrations probably brought additional uranium and other metals to these formations, and the organic matter provided a reducing environment for entrapment. ?? 1990.

Study on Utilization of Super Grade Plutonium in Molten Salt Reactor FUJI-U3 using CITATION Code

NASA Astrophysics Data System (ADS)

Wulandari, Cici; Waris, Abdul; Pramuditya, Syeilendra; Asril, Pramutadi AM; Novitrian

2017-07-01

FUJI-U3 type of Molten Salt Reactor (MSR) has a unique design since it consists of three core regions in order to avoid the replacement of graphite as moderator. MSR uses floride as a nuclear fuel salt with the most popular chemical composition is LiF-BeF2-ThF4-233UF4. ThF4 and 233UF4 are the fertile and fissile materials, respectively. On the other hand, LiF and BeF2 working as both fuel and heat transfer medium. In this study, the super grade plutonium will be utilized as substitution of 233U since plutonium is easier to be obtained compared to 233U as main fuel. Neutronics calculation was performed by using PIJ and CITATION modules of SRAC 2002 code with JENDL 3.2 as nuclear data library.

Initial conceptual design study of self-critical nuclear pumped laser systems

NASA Technical Reports Server (NTRS)

Rodgers, R. J.

1979-01-01

An analytical study of self-critical nuclear pumped laser system concepts was performed. Primary emphasis was placed on reactor concepts employing gaseous uranium hexafluoride (UF6) as the fissionable material. Relationships were developed between the key reactor design parameters including reactor power level, critical mass, neutron flux level, reactor size, operating pressure, and UF6 optical properties. The results were used to select a reference conceptual laser system configuration. In the reference configuration, the 3.2 m cubed lasing volume is surrounded by a graphite internal moderator and a region of heavy water. Results of neutronics calculations yield a critical mass of 4.9 U(235) in the form (235)UF6. The configuration appears capable of operating in a continuous steady-state mode. The average gas temperature in the core is 600 K and the UF6 partial pressure within the lasing volume is 0.34 atm.

One pot synthesis of highly luminescent polyethylene glycol anchored carbon dots functionalized with a nuclear localization signal peptide for cell nucleus imaging.

PubMed

Yang, Lei; Jiang, Weihua; Qiu, Lipeng; Jiang, Xuewei; Zuo, Daiying; Wang, Dongkai; Yang, Li

2015-04-14

Strong blue fluorescent polyethylene glycol (PEG) anchored carbon nitride dots (CDs@PEG) with a high quantum yield (QY) of 75.8% have been synthesized by a one step hydrothermal treatment. CDs with a diameter of ca. 6 nm are well dispersed in water and present a graphite-like structure. Photoluminescence (PL) studies reveal that CDs display excitation-dependent behavior and are stable under various test conditions. Based on the as-prepared CDs, we designed novel cell nucleus targeting imaging carbon dots functionalized with a nuclear localization signal (NLS) peptide. The favourable biocompatibilities of CDs and NLS modified CDs (NLS-CDs) are confirmed by in vitro cytotoxicity assays. Importantly, intracellular localization experiments in MCF7 and A549 cells demonstrate that NLS-CDs could be internalized in the nucleus and show blue light, which indicates that CDs may serve as cell nucleus imaging probes.

Morphological and functional effects of graphene on the synthesis of uranium carbide for isotopes production targets.

PubMed

Biasetto, L; Corradetti, S; Carturan, S; Eloirdi, R; Amador-Celdran, P; Staicu, D; Blanco, O Dieste; Andrighetto, A

2018-05-29

The development of tailored targets for the production of radioactive isotopes represents an active field in nuclear research. Radioactive beams find applications in nuclear medicine, in astrophysics, matter physics and materials science. In this work, we study the use of graphene both as carbon source for UO 2 carbothermal reduction to produce UC x targets, and also as functional properties booster. At fixed composition, the UC x target grain size, porosity and thermal conductivity represent the three main points that affect the target production efficiency. UC x was synthesized using both graphite and graphene as the source of carbon and the target properties in terms of composition, grain size, porosity, thermal diffusivity and thermal conductivity were studied. The main output of this work is related to the remarkable enhancement achieved in thermal conductivity, which can profitably improve thermal dissipation during operational stages of UC x targets.

Portable vibro-acoustic testing system for in situ microstructure characterization and metrology

NASA Astrophysics Data System (ADS)

Smith, James A.; Nichol, Corrie I.; Zuck, Larry D.; Fatemi, Mostafa

2018-04-01

There is a need in research reactors like the one at INL to inspect irradiated materials and structures. The goal of this work is to develop a portable scanning infrastructure for a material characterization technique called vibro-acoustography (VA) that has been developed by the Idaho National laboratory for nuclear applications to characterize fuel, cladding materials, and structures. The proposed VA technology is based on ultrasound and acoustic waves; however, it provides information beyond what is available from the traditional ultrasound techniques and can expand the knowledge on nuclear material characterization and microstructure evolution. This paper will report on the development of a portable scanning system that will be set up to characterize materials and components in open water reactors and canals in situ. We will show some initial laboratory results of images generated by vibro-acoustics of surrogate fuel plates and graphite structures and discuss the design of the portable system.

The advantages and disadvantages of using the TREAT reactor for nuclear laser experiments

NASA Astrophysics Data System (ADS)

Dickson, P. W.; Snyder, A. M.; Imel, G. R.; McConnell, R. J.

The Transient Reactor Test Facility (TREAT) is a large air-cooled test facility located at the Idaho National Engineering Laboratory. Two of the major design features of TREAT, its large size and its being an air-cooled reactor, provide clues to both its advantages and disadvantages for supporting nuclear laser experiments. Its large size, which is dictated by the dilute uranium/graphite fuel, permits accommodation of geometrically large experiments. However, TREAT's large size also results in relatively long transients so that the energy deposited in an experiment is large relative to the peak power available from the reactor. TREAT's air-cooling mode of operation allows its configuration to be changed fairly readily. Due to air cooling, the reactor cools down slowly, permitting only one full power transient a day, which can be a disadvantage in some experimental programs. The reactor is capable of both steady-state or transient operation.

Neutron-rich isotope production using a uranium carbide - carbon nanotubes SPES target prototype

NASA Astrophysics Data System (ADS)

Corradetti, S.; Biasetto, L.; Manzolaro, M.; Scarpa, D.; Carturan, S.; Andrighetto, A.; Prete, G.; Vasquez, J.; Zanonato, P.; Colombo, P.; Jost, C. U.; Stracener, D. W.

2013-05-01

The SPES (Selective Production of Exotic Species) project, under development at the Istituto Nazionale di Fisica Nucleare - Laboratori Nazionali di Legnaro (INFN-LNL), is a new-generation Isotope Separation On-Line (ISOL) facility for the production of radioactive ion beams by means of the proton-induced fission of uranium. In the framework of the research on the SPES target, seven uranium carbide discs, obtained by reacting uranium oxide with graphite and carbon nanotubes, were irradiated with protons at the Holifield Radioactive Ion Beam Facility (HRIBF) of Oak Ridge National Laboratory (ORNL). In the following, the yields of several fission products obtained during the experiment are presented and discussed. The experimental results are then compared to those obtained using a standard uranium carbide target. The reported data highlights the capability of the new type of SPES target to produce and release isotopes of interest for the nuclear physics community.

The early development of neutron diffraction: science in the wings of the Manhattan Project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mason, T. E., E-mail: masont@ornl.gov; Gawne, T. J.; Nagler, S. E.

2013-01-01

Early neutron diffraction experiments performed in 1944 using the first nuclear reactors are described. Although neutron diffraction was first observed using radioactive decay sources shortly after the discovery of the neutron, it was only with the availability of higher intensity neutron beams from the first nuclear reactors, constructed as part of the Manhattan Project, that systematic investigation of Bragg scattering became possible. Remarkably, at a time when the war effort was singularly focused on the development of the atomic bomb, groups working at Oak Ridge and Chicago carried out key measurements and recognized the future utility of neutron diffraction quitemore » independent of its contributions to the measurement of nuclear cross sections. Ernest O. Wollan, Lyle B. Borst and Walter H. Zinn were all able to observe neutron diffraction in 1944 using the X-10 graphite reactor and the CP-3 heavy water reactor. Subsequent work by Wollan and Clifford G. Shull, who joined Wollan’s group at Oak Ridge in 1946, laid the foundations for widespread application of neutron diffraction as an important research tool.« less

Nuclear Thermal Rocket Simulation in NPSS

NASA Technical Reports Server (NTRS)

Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas M.

2013-01-01

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic-metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

Nuclear Thermal Rocket Simulation in NPSS

NASA Technical Reports Server (NTRS)

Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas L.

2013-01-01

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic- metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

Fundamental considerations in dynamic fracture in nuclear materials

NASA Astrophysics Data System (ADS)

Cady, Carl; Eastwood, David; Bourne, Neil; Pei, Ruizhi; Mummery, Paul; Rau, Christoph

2017-06-01

The structural integrity of components used in nuclear power plants is the biggest concern of operators. A diverse range of materials, loading, prior histories and environmental conditions, leads to a complex operating environment. An experimental technique has been developed to characterize brittle materials and using linear elastic fracture mechanics, has given accurate measurements of the fracture toughness of materials. X-ray measurements were used to track the crack front as a function of loading parameters as well as determine the crack surface area as loads increased. This X-ray tomographic study of dynamic fracture in beryllium indicates the onset of damage within the target as load is increased. Similarly, measurements on nuclear graphite were conducted to evaluate the technique. This new, quantitative information obtained using the X-ray techniques has shown application in other materials. These materials exhibited a range of brittle and ductile responses that will test our modelling schemes for fracture. Further visualization of crack front advance and the correlated strain fields that are generated during the experiment for the two distinct deformation processes provide a vital step in validating new multiscale predicative modelling.

ATIC Experiment: Preliminary Results from the Flight in 2002

NASA Technical Reports Server (NTRS)

Ahn, H. S.; Adams, J. H.; Bashindzhagyan, G.; Batkov, K. E.; Chang, J.; Christl, M.; Cox, M.; Ellison, S. B.; Fazely, A. R.; Ganel, O.

2003-01-01

Abstract The Advanced Thin Ionization Calorimeter (ATIC) had successful Long Duration Balloon flights from McMurdo, Antarctica in both 2000 and 2002. The instrument consists of a Silicon matrix for charge measurement, a flared graphite target to induce nuclear interactions, scintillator strip hodoscopes for triggering and helping reconstruct trajectory, and a BGO calorimeter to measure the energy of incident particles. In this paper, we discuss the second flight, which lasted 20 days, starting on 12/29/02. Preliminary results from the on-going analysis of the data including the proton and helium spectra are reported.

ATIC Experiment: Elemental Spectra from the Flight in 2000

NASA Technical Reports Server (NTRS)

Ahn, H. S.; Adams, J. H.; Bashindzhagyan, G.; Batkov, K. E.; Chang, J.; Christl, M.; Fazely, A. R.; Ganel, O.; Gunasingha, R. M.; Guzik, T. G.

2003-01-01

The Advanced Thin Ionization Calorimeter (ATIC) had successful Long Duration Balloon flights from McMurdo, Antarctica in both 2000 and 2002. The instrument consists of a silicon matrix charge detector, a 0.75 nuclear interaction length graphite target, 3 scintillator strip hodoscopes, and an 18 radiation length thick BGO calorimeter to measure the cosmic ray composition and energy spectra from approximately 30 GeV to near 100 TeV. In this paper, we present preliminary results from the first flight, which was a test flight that lasted for 16 days, starting on 12/28/00.

NEUTRONIC REACTOR

DOEpatents

Wigner, E.P.; Weinberg, A.W.; Young, G.J.

1958-04-15

A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

Bridged graphite oxide materials

NASA Technical Reports Server (NTRS)

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.

Preparation of graphitic articles

DOEpatents

Phillips, Jonathan; Nemer, Martin; Weigle, John C.

2010-05-11

Graphitic structures have been prepared by exposing templates (metal, metal-coated ceramic, graphite, for example) to a gaseous mixture that includes hydrocarbons and oxygen. When the template is metal, subsequent acid treatment removes the metal to yield monoliths, hollow graphitic structures, and other products. The shapes of the coated and hollow graphitic structures mimic the shapes of the templates.

Method of Obtaining Uniform Coatings on Graphite

DOEpatents

Campbell, I. E.

1961-04-01

A method is given for obtaining uniform carbide coatings on graphite bodies. According to the invention a metallic halide in vapor form is passed over the graphite body under such conditions of temperature and pressure that the halide reacts with the graphite to form a coating of the metal carbide on the surface of the graphite.

METHOD OF OBTAINING UNIFORM COATINGS ON GRAPHITE

DOEpatents

Campbell, I.E.

1961-04-01

A method is given for obtaining uniform carbide coatings on graphite bodies. According to the invention a metallic halide in vapor form is passed over the graphite body under such conditions of temperature and pressure that the halide reacts with the graphite to form a coating of the metal carbide on the surface of the graphite.

Morphological and optoelectronic characteristics of nanocomposites comprising graphene nanosheets and poly(3-hexylthiophene).

PubMed

Chang, Yo-Wei; Yu, Shiau-Wei; Liu, Cheng-Hao; Tsiang, Raymond Chien-Chao

2010-10-01

P3HT/graphene nanocomposite was prepared via in-situ reduction of exfoliated graphite oxide in the P3HT polymer matrix, where the exfoliated graphite oxide was formed beforehand via the oxidation of graphite via the Hummers method. The oxidation reaction not only imparts functional groups, such as C=O, C-OH, and C-O-C, to graphite but also causes exfoliation of the resulting graphite oxide. The functional groups render graphite oxide an additional, lower thermal degradation temperature (T(d)) and the exfoliation shifts the XRD pattern towards a much smaller angle. The oxidation of graphite into graphite oxide creates a pleated flaking morphology for graphite oxide as opposed to that of graphite. UV/Vis and photoluminescence (PL) spectra of P3HT/graphene nanocomposite indicate that the existence of graphene does not alter the UV/Vis and PL excitation characteristics of P3HT, and the P3HT/graphene composite has higher electron mobility, a smaller band gap and higher conductivity than the pristine P3HT.

Nuclear resonance reflectivity from a [57Fe/Cr]30 multilayer with the Synchrotron Mössbauer Source.

PubMed

Andreeva, Marina A; Baulin, Roman A; Chumakov, Aleksandr I; Rüffer, Rudolf; Smirnov, Gennadii V; Babanov, Yurii A; Devyaterikov, Denis I; Milyaev, Mikhail A; Ponomarev, Dmitrii A; Romashev, Lazar N; Ustinov, Vladimir V

2018-03-01

Mössbauer reflectivity spectra and nuclear resonance reflectivity (NRR) curves have been measured using the Synchrotron Mössbauer Source (SMS) for a [ 57 Fe/Cr] 30 periodic multilayer, characterized by the antiferromagnetic interlayer coupling between adjacent 57 Fe layers. Specific features of the Mössbauer reflectivity spectra measured with π-polarized radiation of the SMS near the critical angle and at the `magnetic' maximum on the NRR curve are analyzed. The variation of the ratio of lines in the Mössbauer reflectivity spectra and the change of the intensity of the `magnetic' maximum under an applied external field has been used to reveal the transformation of the magnetic alignment in the investigated multilayer.

Graphite Sheet Coating for Improved Thermal Oxidative Stability of Carbon Fiber Reinforced/PMR-15 Composites

NASA Technical Reports Server (NTRS)

Campbell, Sandi; Papadopoulos, Demetrios; Heimann, Paula; Inghram, Linda; McCorkle, Linda

2005-01-01

Expanded graphite was compressed into graphite sheets and used as a coating for carbon fiber reinforced PMR-15 composites. BET analysis of the graphite indicated an increase in graphite pore size on compression, however the material was proven to be an effective barrier to oxygen when prepegged with PMR-15 resin. Oxygen permeability of the PMR-15/graphite was an order of magnitude lower than the compressed graphite sheet. By providing a barrier to oxygen permeation, the rate of oxidative degradation of PMR-15 was decreased. As a result, the composite thermo-oxidative stability increased by up to 25%. The addition of a graphite sheet as a top ply on the composites yielded little change in the material's flexural strength or interlaminar shear strength.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marquez, Eva; Pina, Gabriel; Rodriguez, Marina

Spain has to manage about 3700 tons of irradiated graphite from the reactor Vandellos I as radioactive waste. 2700 tons are the stack of the reactor and are still in the reactor core waiting for retrieval. The rest of the quantities, 1000 tons, are the graphite sleeves which have been already retrieved from the reactor. During operation the graphite sleeves were stored in a silo and during the dismantling stage a retrieval process was carried out separating the wires from the graphite, which were crushed and introduced into 220 cubic containers of 6 m{sup 3} each and placed in interimmore » storage. The graphite is an intermediate level radioactive waste but it contains long lived radionuclides like {sup 14}C which disqualifies disposal at the low level waste repository of El Cabril. Therefore, a new project has been started in order to investigate two new options for the management of this waste type. The first one is based on a selective decontamination of {sup 14}C by thermal methods. This method is based on results obtained at the Research Centre Juelich (FZJ) in the Frame of the EC programs 'Raphael' and 'Carbowaste'. The process developed at FZJ is based on a preferential oxidation of {sup 14}C in comparison to the bulk {sup 12}C. Explanations for this effect are the inhomogeneous distribution and a weaker bounding of {sup 14}C which is not incorporated in the graphite lattice. However these investigations have only been performed with graphite from the high temperature reactor Arbeitsgemeinschaft Versuchsreaktor Juelich AVR which has been operated in a non-oxidising condition or research reactor graphite operated at room temperature. The reactor Vandellos I has been operated with CO{sub 2} as coolant and significant amounts of graphite have been already oxidised. The aim of the project is to validate whether a {sup 14}C decontamination can also been achieved with graphite from Vandellos I. A second possibility under investigation is the encapsulation of the graphite in a long term stable glass matrix. The principal applicability has been already proved by FNAG. Crushed graphite mixed with a suitable glass powder has been pressed at elevated temperature under vacuum. The vacuum is required to avoid gas enclosures in the obtained product. The obtained products, named IGM for 'Impermeable Graphite Matrix', have densities above 99% of theoretical density. The amount of glass has been chosen with respect to the pore volume of the former graphite parts. The method allows the production of encapsulated graphite without increasing the disposal volume. This paper will give a short overview of characterisation results of different irradiated graphite materials obtained at CIEMAT and in the Carbowaste project as well as the proposed methods and the actual status of the program including first results about leaching of non-radioactive IGM samples and hopefully first tendencies concerning the C-14 separation from graphite of Vandellos I by thermal treatment. Both processes, the thermal treatment as well as the IGM, have the potential to solve problems related to the management of irradiated graphite in Spain. However the methods have only been tested with different types of i-graphite and virgin graphite, respectively. Only investigations with real i-graphite from Spain will reveal whether the described methods are applicable to graphite from Vandellos I. However all partners are convinced that one of these new methods or a combination of them will lead to a feasible option to manage i-graphite in Spain on an industrial scale. (authors)« less

Research on graphite reinforced glass matrix composites

NASA Technical Reports Server (NTRS)

Bacon, J. F.; Prewo, K. M.

1977-01-01

The results of research for the origination of graphite-fiber reinforced glass matrix composites are presented. The method selected to form the composites consisted of pulling the graphite fiber through a slurry containing powdered glass, winding up the graphite fiber and the glass it picks up on a drum, drying, cutting into segments, loading the tape segment into a graphite die, and hot pressing. During the course of the work, composites were made with a variety of graphite fibers in a glass matrix.

Energy shadowing correction of ultrasonic pulse-echo records by digital signal processing

NASA Technical Reports Server (NTRS)

Kishoni, D.; Heyman, J. S.

1986-01-01

Attention is given to a numerical algorithm that, via signal processing, enables the dynamic correction of the shadowing effect of reflections on ultrasonic displays. The algorithm was applied to experimental data from graphite-epoxy composite material immersed in a water bath. It is concluded that images of material defects with the shadowing corrections allow for a more quantitative interpretation of the material state. It is noted that the proposed algorithm is fast and simple enough to be adopted for real time applications in industry.

Research on the transformation mechanism of graphite phase and microstructure in the heated region of gray cast iron by laser cladding

NASA Astrophysics Data System (ADS)

Liu, Yancong; Zhan, Xianghua; Yi, Peng; Liu, Tuo; Liu, Benliang; Wu, Qiong

2018-03-01

A double-track lap cladding experiment involving gray cast iron was established to investigate the transformation mechanism of graphite phase and microstructure in a laser cladding heated region. The graphite phase and microstructure in different heated regions were observed under a microscope, and the distribution of elements in various heated regions was analyzed using an electron probe. Results show that no graphite existed in the cladding layer and in the middle and upper parts of the binding region. Only some of the undissolved small graphite were observed at the bottom of the binding region. Except the refined graphite size, the morphological characteristics of substrate graphite and graphite in the heat-affected zone were similar. Some eutectic clusters, which grew along the direction of heat flux, were observed in the heat-affected zone whose microstructure was transformed into a mixture of austenite, needle-like martensite, and flake graphite. Needle-like martensite around graphite was fine, but this martensite became sparse and coarse when it was away from graphite. Some martensite clusters appeared in the local area near the binding region, and the carbon atoms in the substrate did not diffuse into the cladding layer through laser cladding, which only affected the bonding area and the bottom of the cladding layer.

Nickel Nanoparticle Encapsulated in Few-Layer Nitrogen-Doped Graphene Supported by Nitrogen-Doped Graphite Sheets as a High-Performance Electromagnetic Wave Absorbing Material.

PubMed

Yuan, Haoran; Yan, Feng; Li, Chunyan; Zhu, Chunling; Zhang, Xitian; Chen, Yujin

2018-01-10

Herein we develop a facile strategy for fabricating nickel particle encapsulated in few-layer nitrogen-doped graphene supported by graphite carbon sheets as a high-performance electromagnetic wave (EMW) absorbing material. The obtained material exhibits sheetlike morphology with a lateral length ranging from a hundred nanometers to 2 μm and a thickness of about 23 nm. Nickel nanoparticles with a diameter of approximately 20 nm were encapsulated in about six layers of nitrogen-doped graphene. As applied for electromagnetic absorbing material, the heteronanostructures exhibit excellent electromagnetic wave absorption property, comparable to most EMW absorbing materials previously reported. Typically, the effective absorption bandwidth (the frequency region falls within the reflection loss below -10 dB) is up to 8.5 GHz at the thicknesses of 3.0 mm for the heteronanostructures with the optimized Ni content. Furthermore, two processes, carbonization at a high temperature and subsequent treatment in hot acid solution, were involved in the preparation of the heteronanostructures, and thus, mass production was achieved easily, facilitating their practical applications.

Synthesis and characterization of TiO2/graphitic carbon nanocomposites with enhanced photocatalytic performance

NASA Astrophysics Data System (ADS)

Wanag, Agnieszka; Kusiak-Nejman, Ewelina; Kowalczyk, Łukasz; Kapica-Kozar, Joanna; Ohtani, Bunsho; Morawski, Antoni W.

2018-04-01

In this paper titanium dioxide carbon modification with benzene as a carbon source is presented. A TiO2/graphitic carbon nanocomposites were synthesized by thermal modification in the presence of benzene vapours at different temperature (300-700 °C). The new materials were characterized by a various techniques, such as: X-ray diffraction (XRD), scanning electron microscopy (SEM), diffuse reflectance spectroscopy (UV-vis/DR), surface-enhanced Raman spectroscopy. BET specific surface area was also measured. The photocatalytic activity of obtained nanocomposites was measured by the decomposition of acetic acid and methylene blue under UV-vis irradiation. The results show that photocatalytic activity increasing with increase in carbon concentration and temperature of modification. It can be noted that adsorption degree has a very high impact on methylene blue decomposition. The highest photocatalytic activity was found for the photocatalyst modified at 600 °C contains 1.13 wt% of carbon. It should be noted that, the influence of crystallite size, crystal structure changes and specific surface area for photocatalytic activity are presented.

Laser shock wave assisted patterning on NiTi shape memory alloy surfaces

NASA Astrophysics Data System (ADS)

Seyitliyev, Dovletgeldi; Li, Peizhen; Kholikov, Khomidkhodza; Grant, Byron; Karaca, Haluk E.; Er, Ali O.

2017-02-01

An advanced direct imprinting method with low cost, quick, and less environmental impact to create thermally controllable surface pattern using the laser pulses is reported. Patterned micro indents were generated on Ni50Ti50 shape memory alloys (SMA) using an Nd:YAG laser operating at 1064 nm combined with suitable transparent overlay, a sacrificial layer of graphite, and copper grid. Laser pulses at different energy densities which generates pressure pulses up to 10 GPa on the surface was focused through the confinement medium, ablating the copper grid to create plasma and transferring the grid pattern onto the NiTi surface. Scanning electron microscope (SEM) and optical microscope images of square pattern with different sizes were studied. One dimensional profile analysis shows that the depth of the patterned sample initially increase linearly with the laser energy until 125 mJ/pulse where the plasma further absorbs and reflects the laser beam. In addition, light the microscope image show that the surface of NiTi alloy was damaged due to the high power laser energy which removes the graphite layer.

NEW METHOD OF GRAPHITE PREPARATION

DOEpatents

Stoddard, S.D.; Harper, W.T.

1961-08-29

BS>A method is described for producing graphite objects comprising mixing coal tar pitch, carbon black, and a material selected from the class comprising raw coke, calcined coke, and graphite flour. The mixture is placed in a graphite mold, pressurized to at least 1200 psi, and baked and graphitized by heating to about 2500 deg C while maintaining such pressure. (AEC)

Monte Carlo Calculation of Thermal Neutron Inelastic Scattering Cross Section Uncertainties by Sampling Perturbed Phonon Spectra

NASA Astrophysics Data System (ADS)

Holmes, Jesse Curtis

Nuclear data libraries provide fundamental reaction information required by nuclear system simulation codes. The inclusion of data covariances in these libraries allows the user to assess uncertainties in system response parameters as a function of uncertainties in the nuclear data. Formats and procedures are currently established for representing covariances for various types of reaction data in ENDF libraries. This covariance data is typically generated utilizing experimental measurements and empirical models, consistent with the method of parent data production. However, ENDF File 7 thermal neutron scattering library data is, by convention, produced theoretically through fundamental scattering physics model calculations. Currently, there is no published covariance data for ENDF File 7 thermal libraries. Furthermore, no accepted methodology exists for quantifying or representing uncertainty information associated with this thermal library data. The quality of thermal neutron inelastic scattering cross section data can be of high importance in reactor analysis and criticality safety applications. These cross sections depend on the material's structure and dynamics. The double-differential scattering law, S(alpha, beta), tabulated in ENDF File 7 libraries contains this information. For crystalline solids, S(alpha, beta) is primarily a function of the material's phonon density of states (DOS). Published ENDF File 7 libraries are commonly produced by calculation and processing codes, such as the LEAPR module of NJOY, which utilize the phonon DOS as the fundamental input for inelastic scattering calculations to directly output an S(alpha, beta) matrix. To determine covariances for the S(alpha, beta) data generated by this process, information about uncertainties in the DOS is required. The phonon DOS may be viewed as a probability density function of atomic vibrational energy states that exist in a material. Probable variation in the shape of this spectrum may be established that depends on uncertainties in the physics models and methodology employed to produce the DOS. Through Monte Carlo sampling of perturbations from the reference phonon spectrum, an S(alpha, beta) covariance matrix may be generated. In this work, density functional theory and lattice dynamics in the harmonic approximation are used to calculate the phonon DOS for hexagonal crystalline graphite. This form of graphite is used as an example material for the purpose of demonstrating procedures for analyzing, calculating and processing thermal neutron inelastic scattering uncertainty information. Several sources of uncertainty in thermal neutron inelastic scattering calculations are examined, including sources which cannot be directly characterized through a description of the phonon DOS uncertainty, and their impacts are evaluated. Covariances for hexagonal crystalline graphite S(alpha, beta) data are quantified by coupling the standard methodology of LEAPR with a Monte Carlo sampling process. The mechanics of efficiently representing and processing this covariance information is also examined. Finally, with appropriate sensitivity information, it is shown that an S(alpha, beta) covariance matrix can be propagated to generate covariance data for integrated cross sections, secondary energy distributions, and coupled energy-angle distributions. This approach enables a complete description of thermal neutron inelastic scattering cross section uncertainties which may be employed to improve the simulation of nuclear systems.

Electronic and total energy properties of ternary and quaternary semiconductor compounds, alloys, and superlattices: Theoretical study of Cu/graphite bonding

NASA Technical Reports Server (NTRS)

Lambrecht, Walter R. L.

1992-01-01

The goals of the research were to provide a fundamental science basis for why the bonding of Cu to graphite is weak, to critically evaluate the previous analysis of the wetting studies with particular regard to the values used for the surface energies of Cu and graphite, and to make recommendations for future experiments or other studies which could advance the understanding and solution of this technological problem. First principles electronic structure calculations were used to study the problem. These are based on density functional theory in the local density approximation and the use of the linear muffin-tin orbital band structure method. Calculations were performed for graphite monolayers, single crystal graphite with the hexagonal AB stacking, bulk Cu, Cu(111) surface, and Cu/graphite superlattices. The study is limited to the basal plane of graphite because this is the graphite plane exposed to Cu and graphite surface energies and combined with the measured contact angles to evaluate the experimental adhesion energy.

Low-energy electron diffraction study of potassium adsorbed on single-crystal graphite and highly oriented pyrolytic graphite

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ferralis, N.; Diehl, R.D.; Pussi, K.

2004-12-15

Potassium adsorption on graphite has been a model system for the understanding of the interaction of alkali metals with surfaces. The geometries of the (2x2) structure of potassium on both single-crystal graphite (SCG) and highly oriented pyrolytic graphite (HOPG) were investigated for various preparation conditions for graphite temperatures between 55 and 140 K. In all cases, the geometry was found to consist of K atoms in the hollow sites on top of the surface. The K-graphite average perpendicular spacing is 2.79{+-}0.03 A , corresponding to an average C-K distance of 3.13{+-}0.03 A , and the spacing between graphite planes ismore » consistent with the bulk spacing of 3.35 A. No evidence was observed for a sublayer of potassium. The results of dynamical LEED studies for the clean SCG and HOPG surfaces indicate that the surface structures of both are consistent with the truncated bulk structure of graphite.« less

An Electron Microscopy Study of Graphite Growth in Nodular Cast Irons

NASA Astrophysics Data System (ADS)

Laffont, L.; Jday, R.; Lacaze, J.

2018-04-01

Growth of graphite during solidification and high-temperature solid-state transformation has been investigated in samples cut out from a thin-wall casting which solidified partly in the stable (iron-graphite) and partly in the metastable (iron-cementite) systems. Transmission electron microscopy has been used to characterize graphite nodules in as-cast state and in samples having been fully graphitized at various temperatures in the austenite field. Nodules in the as-cast material show a twofold structure characterized by an inner zone where graphite is disoriented and an outer zone where it is well crystallized. In heat-treated samples, graphite nodules consist of well-crystallized sectors radiating from the nucleus. These observations suggest that the disoriented zone appears because of mechanical deformation when the liquid contracts during its solidification in the metastable system. During heat-treatment, the graphite in this zone recrystallizes. In turn, it can be concluded that nodular graphite growth mechanism is the same during solidification and solid-state transformation.

Understanding the crack formation of graphite particles in cycled commercial lithium-ion batteries by focused ion beam - scanning electron microscopy

NASA Astrophysics Data System (ADS)

Lin, Na; Jia, Zhe; Wang, Zhihui; Zhao, Hui; Ai, Guo; Song, Xiangyun; Bai, Ying; Battaglia, Vincent; Sun, Chengdong; Qiao, Juan; Wu, Kai; Liu, Gao

2017-10-01

The structure degradation of commercial Lithium-ion battery (LIB) graphite anodes with different cycling numbers and charge rates was investigated by focused ion beam (FIB) and scanning electron microscopy (SEM). The cross-section image of graphite anode by FIB milling shows that cracks, resulted in the volume expansion of graphite electrode during long-term cycling, were formed in parallel with the current collector. The crack occurs in the bulk of graphite particles near the lithium insertion surface, which might derive from the stress induced during lithiation and de-lithiation cycles. Subsequently, crack takes place along grain boundaries of the polycrystalline graphite, but only in the direction parallel with the current collector. Furthermore, fast charge graphite electrodes are more prone to form cracks since the tensile strength of graphite is more likely to be surpassed at higher charge rates. Therefore, for LIBs long-term or high charge rate applications, the tensile strength of graphite anode should be taken into account.

Nuclear Strategy and World Order: The United States Imperative.

ERIC Educational Resources Information Center

Beres, Louis Rene

The current U.S. nuclear strategy goes beyond the legitimate objective of survivable strategic forces to active preparation for nuclear war. The Reagan administration strategy rejects minimum deterrence and prepares for a nuclear war that might be protracted and controlled. The strategy reflects the understanding that a combination of counterforce…

Friction and wear of carbon-graphite materials for high-energy brakes

NASA Technical Reports Server (NTRS)

Bill, R. C.

1978-01-01

Caliper type brake simulation experiments were conducted on seven different carbon graphite materials formulations against a steel disk material and against a carbon graphite disk material. The effects of binder level, boron carbide (B4C) additions, SiC additions, graphite fiber additions, and graphite cloth reinforcement on friction and wear behavior were investigated. Reductions in binder level, additions of B4C, and additions of SiC each resulted in increased wear. The wear rate was not affected by the addition of graphite fibers. Transition to severe wear and high friction was observed in the case of graphite-cloth-reinforced carbon sliding against a disk of similar composition. The transition was related to the disruption of a continuous graphite shear film that must form on the sliding surfaces if low wear is to occur.

AGR-3/4 Irradiation Test Train Disassembly and Component Metrology First Look Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stempien, John Dennis; Rice, Francine Joyce; Harp, Jason Michael

2016-03-01

The AGR-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing UCO TRISO particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer which was intended to fail during irradiation and provide a source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear graphite. Through post-irradiation examination (PIE) of themore » rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g. diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. No corresponding discoloration was found on the inside of these capsules, however. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools (including clamps, cutters, and drills) had been designed and fabricated in order to carry out test train disassembly and recovery of capsule components (graphite rings and fuel compacts). This equipment performed well for separating each capsule in the test train and extracting the capsule components. Only a few problems were encountered. In one case, the outermost ring (the sink ring) was cracked during removal of the capsule through tubes. Although the sink ring will be analyzed in order to obtain a mass balance of fission products in the experiment, these cracks do not pose a major concern because the sink ring will not be analyzed in detail to obtain the spatial distribution of fission products. In Capsules 4 and 5, the compacts could not be removed from the inner rings. Strategies for removing the compacts are being evaluated. Sampling the inner rings with the compacts in-place is also an option. Dimensional measurements were made on the compacts, inner rings, outer rings, and sink rings. The diameters of all compacts decreased by 0.5 to 2.0 %. Generally, the extent of diametric shrinkage increased linearly with increasing neutron fluence. Most compact lengths also decreased. Compact lengths decreased with increasing fluence, reaching maximum shrinkage of about 0.9 % at a fast fluence of 4.0x10 25 n/m 2 E > 0.18 MeV. Above this fluence, the extent of length shrinkage appeared to decrease with fluence, and two compacts from Capsule 7 were found to have slightly increased in length (< 0.1 %) after a fluence of 5.2x10 25 n/m 2.« less

AGR-3/4 Irradiation Test Train Disassembly and Component Metrology First Look Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stempien, John Dennis; Rice, Francine Joyce; Harp, Jason Michael

The AGR-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing UCO TRISO particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer which was intended to fail during irradiation and provide a source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear graphite. Through post-irradiation examination (PIE) of themore » rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g. diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. No corresponding discoloration was found on the inside of these capsules, however. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools (including clamps, cutters, and drills) had been designed and fabricated in order to carry out test train disassembly and recovery of capsule components (graphite rings and fuel compacts). This equipment performed well for separating each capsule in the test train and extracting the capsule components. Only a few problems were encountered. In one case, the outermost ring (the sink ring) was cracked during removal of the capsule through tubes. Although the sink ring will be analyzed in order to obtain a mass balance of fission products in the experiment, these cracks do not pose a major concern because the sink ring will not be analyzed in detail to obtain the spatial distribution of fission products. In Capsules 4 and 5, the compacts could not be removed from the inner rings. Strategies for removing the compacts are being evaluated. Sampling the inner rings with the compacts in-place is also an option. Dimensional measurements were made on the compacts, inner rings, outer rings, and sink rings. The diameters of all compacts decreased by 0.5 to 2.0 %. Generally, the extent of diametric shrinkage increased linearly with increasing neutron fluence. Most compact lengths also decreased. Compact lengths decreased with increasing fluence, reaching maximum shrinkage of about 0.9 % at a fast fluence of 4.0x1025 n/m2 E > 0.18 MeV. Above this fluence, the extent of length shrinkage appeared to decrease with fluence, and two compacts from Capsule 7 were found to have slightly increased in length (< 0.1 %) after a fluence of 5.2x1025 n/m2.« less

Tubular graphite cones.

PubMed

Zhang, Guangyu; Jiang, Xin; Wang, Enge

2003-04-18

We report the synthesis of tubular graphite cones using a chemical vapor deposition method. The cones have nanometer-sized tips, micrometer-sized roots, and hollow interiors with a diameter ranging from about 2 to several tens of nanometers. The cones are composed of cylindrical graphite sheets; a continuous shortening of the graphite layers from the interior to the exterior makes them cone-shaped. All of the tubular graphite cones have a faceted morphology. The constituent graphite sheets have identical chiralities of a zigzag type across the entire diameter, imparting structural control to tubular-based carbon structures. The tubular graphite cones have potential for use as tips for scanning probe microscopy, but with greater rigidity and easier mounting than currently used carbon nanotubes.

Process for the fabrication of aluminum metallized pyrolytic graphite sputtering targets

DOEpatents

Makowiecki, D.M.; Ramsey, P.B.; Juntz, R.S.

1995-07-04

An improved method is disclosed for fabricating pyrolytic graphite sputtering targets with superior heat transfer ability, longer life, and maximum energy transmission. Anisotropic pyrolytic graphite is contoured and/or segmented to match the erosion profile of the sputter target and then oriented such that the graphite`s high thermal conductivity planes are in maximum contact with a thermally conductive metal backing. The graphite contact surface is metallized, using high rate physical vapor deposition (HRPVD), with an aluminum coating and the thermally conductive metal backing is joined to the metallized graphite target by one of four low-temperature bonding methods; liquid-metal casting, powder metallurgy compaction, eutectic brazing, and laser welding. 11 figs.

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

DOEpatents

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.

Mineral resource of the month: graphite

USGS Publications Warehouse

,

2008-01-01

The article presents facts about graphite ideal for industrial applications. Among the characteristics of graphite are its metallic luster, softness, perfect basal cleavage and electrical conductivity. Batteries, brake linings and powdered metals are some of the products that make use of graphite. It attributes the potential applications for graphite in high-technology fields to innovations in thermal technology and acid-leaching techniques.

Design modification for the modular helium reactor for higher temperature operation and reliability studies for nuclear hydrogen production processes

NASA Astrophysics Data System (ADS)

Reza, S. M. Mohsin

Design options have been evaluated for the Modular Helium Reactor (MHR) for higher temperature operation. An alternative configuration for the MHR coolant inlet flow path is developed to reduce the peak vessel temperature (PVT). The coolant inlet path is shifted from the annular path between reactor core barrel and vessel wall through the permanent side reflector (PSR). The number and dimensions of coolant holes are varied to optimize the pressure drop, the inlet velocity, and the percentage of graphite removed from the PSR to create this inlet path. With the removal of ˜10% of the graphite from PSR the PVT is reduced from 541°C to 421°C. A new design for the graphite block core has been evaluated and optimized to reduce the inlet coolant temperature with the aim of further reduction of PVT. The dimensions and number of fuel rods and coolant holes, and the triangular pitch have been changed and optimized. Different packing fractions for the new core design have been used to conserve the number of fuel particles. Thermal properties for the fuel elements are calculated and incorporated into these analyses. The inlet temperature, mass flow and bypass flow are optimized to limit the peak fuel temperature (PFT) within an acceptable range. Using both of these modifications together, the PVT is reduced to ˜350°C while keeping the outlet temperature at 950°C and maintaining the PFT within acceptable limits. The vessel and fuel temperatures during low pressure conduction cooldown and high pressure conduction cooldown transients are found to be well below the design limits. The reliability and availability studies for coupled nuclear hydrogen production processes based on the sulfur iodine thermochemical process and high temperature electrolysis process have been accomplished. The fault tree models for both these processes are developed. Using information obtained on system configuration, component failure probability, component repair time and system operating modes and conditions, the system reliability and availability are assessed. Required redundancies are made to improve system reliability and to optimize the plant design for economic performance. The failure rates and outage factors of both processes are found to be well below the maximum acceptable range.

Nucleation and Growth of Graphite in Eutectic Spheroidal Cast Iron: Modeling and Testing

NASA Astrophysics Data System (ADS)

Carazo, Fernando D.; Dardati, Patricia M.; Celentano, Diego J.; Godoy, Luis A.

2016-06-01

A new model of graphite growth during the continuous cooling of eutectic spheroidal cast iron is presented in this paper. The model considers the nucleation and growth of graphite from pouring to room temperature. The microstructural model of solidification accounts for the eutectic as divorced and graphite growth rate as a function of carbon gradient at the liquid in contact with the graphite. In the solid state, the microstructural model takes into account three stages for graphite growth, namely (1) from the end of solidification to the upper bound of intercritical stable eutectoid, (2) during the intercritical stable eutectoid, and (3) from the lower bound of intercritical stable eutectoid to room temperature. The micro- and macrostructural models are coupled using a sequential multiscale approach. Numerical results for graphite fraction and size distribution are compared with experimental results obtained from a cylindrical cup, in which the graphite volumetric fraction and size distribution were obtained using the Schwartz-Saltykov approach. The agreements between the experimental and numerical results for the fraction of graphite and the size distribution of spheroids reveal the importance of numerical models in the prediction of the main aspects of graphite in spheroidal cast iron.

Graphitized-carbon fiber/carbon char fuel

DOEpatents

Cooper, John F [Oakland, CA

2007-08-28

A method for recovery of intact graphitic fibers from fiber/polymer composites is described. The method comprises first pyrolyzing the graphite fiber/polymer composite mixture and then separating the graphite fibers by molten salt electrochemical oxidation.

Phase Transition of H 2 in Subnanometer Pores Observed at 75 K

DOE PAGES

Olsen, Raina J.; Gillespie, Andrew K.; Contescu, Cristian I.; ...

2017-10-30

In this paper, we report a phase transition in H 2 adsorbed in a locally graphitic Saran carbon with subnanometer pores 0.5–0.65 nm in width, in which two layers of hydrogen can just barely squeeze, provided they pack tightly. The phase transition is observed at 75 K, temperatures far higher than other systems in which an adsorbent is known to increase phase transition temperatures: for instance, H 2 melts at 14 K in the bulk, but at 20 K on graphite because the solid H 2 is stabilized by the surface structure. Here we observe a transition at 75 Kmore » and 77–200 bar: from a low-temperature, low-density phase to a high-temperature, higher density phase. We model the low-density phase as a monolayer commensurate solid composed mostly of para-H 2 (the ground nuclear spin state, S = 0) and the high-density phase as an orientationally ordered bilayer commensurate solid composed mostly of ortho-H 2 (S = 1). We attribute the increase in density with temperature to the fact that the oblong ortho-H 2 can pack more densely. The transition is observed using two experiments. The high-density phase is associated with an increase in neutron backscatter by a factor of 7.0 ± 0.1. Normally, hydrogen produces no backscatter (scattering angle >90°). This backscatter appears along with a discontinuous increase in the excitation mass from 1.2 amu to 21.0 ± 2.3 amu, which we associate with collective nuclear spin excitations in the orientationally ordered phase. Film densities were measured using hydrogen adsorption. Finally, no phase transition was observed in H 2 adsorbed in control activated carbon materials.« less

Phase Transition of H 2 in Subnanometer Pores Observed at 75 K

DOE Office of Scientific and Technical Information (OSTI.GOV)

Olsen, Raina J.; Gillespie, Andrew K.; Contescu, Cristian I.

In this paper, we report a phase transition in H 2 adsorbed in a locally graphitic Saran carbon with subnanometer pores 0.5–0.65 nm in width, in which two layers of hydrogen can just barely squeeze, provided they pack tightly. The phase transition is observed at 75 K, temperatures far higher than other systems in which an adsorbent is known to increase phase transition temperatures: for instance, H 2 melts at 14 K in the bulk, but at 20 K on graphite because the solid H 2 is stabilized by the surface structure. Here we observe a transition at 75 Kmore » and 77–200 bar: from a low-temperature, low-density phase to a high-temperature, higher density phase. We model the low-density phase as a monolayer commensurate solid composed mostly of para-H 2 (the ground nuclear spin state, S = 0) and the high-density phase as an orientationally ordered bilayer commensurate solid composed mostly of ortho-H 2 (S = 1). We attribute the increase in density with temperature to the fact that the oblong ortho-H 2 can pack more densely. The transition is observed using two experiments. The high-density phase is associated with an increase in neutron backscatter by a factor of 7.0 ± 0.1. Normally, hydrogen produces no backscatter (scattering angle >90°). This backscatter appears along with a discontinuous increase in the excitation mass from 1.2 amu to 21.0 ± 2.3 amu, which we associate with collective nuclear spin excitations in the orientationally ordered phase. Film densities were measured using hydrogen adsorption. Finally, no phase transition was observed in H 2 adsorbed in control activated carbon materials.« less

Characterization of Mechanical Properties of Nuclear Graphite Using Subsize Specimens and Reusing Tested Specimens

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ji Hyun, Yoon; Byun, Thak Sang; Strizak, Joe P

2011-01-01

The mechanical properties of NBG-18 nuclear grade graphite have been characterized using small specimen test techniques and statistical treatment on the test results. New fracture strength and toughness test techniques were developed to use subsize cylindrical specimens with glued heads and to reuse their broken halves. Three sets of subsize cylindrical specimens with the different diameters of 4 mm, 8 mm, and 12 mm were tested to obtain tensile fracture strength. The longer piece of the broken halves was cracked from side surfaces and tested under three-point bend loading to obtain fracture toughness. Both the strength and fracture toughness datamore » were analyzed using Weibull distribution models focusing on size effect. The mean fracture strength decreased from 22.9 MPa to 21.5 MPa as the diameter increased from 4 mm to 12 mm, and the mean strength of 15.9 mm diameter standard specimen, 20.9 MPa, was on the extended trend line. These fracture strength data indicate that in the given diameter range the size effect is not significant and much smaller than that predicted by the Weibull statistics-based model. Further, no noticeable size effect existed in the fracture toughness data, whose mean values were in a narrow range of 1.21 1.26 MPa. The Weibull moduli measured for fracture strength and fracture toughness datasets were around 10. It is therefore believed that the small or negligible size effect enables to use the subsize specimens and that the new fracture toughness test method to reuse the broken specimens to help minimize irradiation space and radioactive waste.« less

Electrochemical treatment of evaporated residue of soak liquor generated from leather industry.

PubMed

Boopathy, R; Sekaran, G

2013-09-15

The organic and suspended solids present in soak liquor, generated from leather industry, demands treatment. The soak liquor is being segregated and evaporated in solar evaporation pans/multiple effect evaporator due to non availability of viable technology for its treatment. The residue left behind in the pans/evaporator does not carry any reuse value and also faces disposal threat due to the presence of high concentration of sodium chloride, organic and bacterial impurities. In the present investigation, the aqueous evaporated residue of soak liquor (ERSL) was treated by electrochemical oxidation. Graphite/graphite and SS304/graphite systems were used in electrochemical oxidation of organics in ERSL. Among these, graphite/graphite system was found to be effective over SS304/graphite system. Hence, the optimised conditions for the electrochemical oxidation of organics in ERSL using graphite/graphite system was evaluated by response surface methodology (RSM). The mass transport coefficient (km) was calculated based on pseudo-first order rate kinetics for both the electrode systems (graphite/graphite and SS304/graphite). The thermodynamic properties illustrated the electrochemical oxidation was exothermic and non-spontaneous in nature. The calculated specific energy consumption at the optimum current density of 50 mA cm(-2) was 0.41 kWh m(-3) for the removal of COD and 2.57 kWh m(-3) for the removal of TKN. Copyright © 2013 Elsevier B.V. All rights reserved.

Influence of oxygen on the carbide formation on tungsten

NASA Astrophysics Data System (ADS)

Luthin, J.; Linsmeier, Ch.

2001-03-01

As a first wall material in nuclear fusion devices, tungsten will interact with carbon and oxygen from the plasma. In this study, we report on the process of thermally induced carbide formation of thin carbon films on polycrystalline tungsten and the influence of oxygen on this process. All investigations are performed using X-ray photoelectron spectroscopy (XPS). Carbon films are supplied through electron beam evaporation of graphite. The carbidization process, monitored during increased substrate temperature, can be divided into four phases. In phase I disordered carbon converts into graphite-like carbon. In phase II significant diffusion and the reaction to W 2C is observed, followed by phase III which is dominated by the presence of W 2C and the beginning reaction to WC. Finally in phase IV only WC is present, but the total carbon amount has strongly decreased. Different mechanisms of oxygen influence on the carbide formation are proposed and measurements of the reaction of carbon on tungsten with intermediate oxide layers are presented in detail. A WO 2+ x intermediate layer completely inhibits the carbide formation, while a WO 2 layer leads to WC formation at temperatures above 1270 K.

Preparation and Characterization of Graphite Waste/CeO2 Composites

NASA Astrophysics Data System (ADS)

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.

Monolithic porous graphitic carbons obtained through catalytic graphitization of carbon xerogels

NASA Astrophysics Data System (ADS)

Kiciński, Wojciech; Norek, Małgorzata; Bystrzejewski, Michał

2013-01-01

Pyrolysis of organic xerogels accompanied by catalytic graphitization and followed by selective-combustion purification was used to produce porous graphitic carbons. Organic gels impregnated with iron(III) chloride or nickel(II) acetate were obtained through polymerization of resorcinol and furfural. During the pyrolysis stage graphitization of the gel matrix occurs, which in turn develops mesoporosity of the obtained carbons. The evolution of the carbon into graphitic structures is strongly dependent on the concentrations of the transition metal. Pyrolysis leads to monoliths of carbon xerogel characterized by substantially enhanced mesoporosity resulting in specific surface areas up to 400 m2/g. Removal of the amorphous carbon by selective-combustion purification reduces the xerogels' mesoporosity, occasionally causing loss of their mechanical strength. The graphitized carbon xerogels were investigated by means of SEM, XRD, Raman scattering, TG-DTA and N2 physisorption. Through this procedure well graphitized carbonaceous materials can be obtained as bulk pieces.

Friction and wear of carbon-graphite materials for high energy brakes

NASA Technical Reports Server (NTRS)

Bill, R. C.

1975-01-01

Caliper-type brakes simulation experiments were conducted on seven different carbon-graphite material formulations against a steel disk material and against a carbon-graphite disk material. The effects of binder level, boron carbide (B4C) additions, graphite fiber additions, and graphite cloth reinforcement on friction and wear behavior were investigated. Reductions in binder level and additions of B4C each resulted in increased wear. The wear rate was not affected by the addition of graphite fibers. Transition to severe wear and high friction was observed in the case of graphite-cloth-reinforced carbon sliding against a disk of similar composition. This transition was related to the disruption of a continuous graphite shear film that must form on the sliding surfaces if low wear is to occur. The exposure of the fiber structure of the cloth constituent is believed to play a role in the shear film disruption.

A preliminary study on the use of (10)Be in forensic radioecology of nuclear explosion sites.

PubMed

Whitehead, N E; Endo, S; Tanaka, K; Takatsuji, T; Hoshi, M; Fukutani, S; Ditchburn, R G; Zondervan, A

2008-02-01

Cosmogenic (10)Be, known for use in dating studies, unexpectedly is also produced in nuclear explosions with an atom yield almost comparable to (e.g.) (137)Cs. There are major production routes via (13)C(n, alpha)(10)Be, from carbon dioxide in the air and the organic explosives, possibly from other bomb components and to a minor extent from the direct fission reaction. Although the detailed bomb components are speculative, carbon was certainly present in the explosives and an order of magnitude calculation is possible. The (n, alpha) cross-section was determined by irradiating graphite in a nuclear reactor, and the resulting (10)Be estimated by Accelerator Mass Spectrometry (AMS) giving a cross-section of 34.5+/-0.7mb (6-9.3MeV), within error of previous work. (10)Be should have applications in forensic radioecology. Historical environmental samples from Hiroshima, and Semipalatinsk (Kazakhstan) showed two to threefold (10)Be excesses compared with the background cosmogenic levels. A sample from Lake Chagan (a Soviet nuclear cratering experiment) contained more (10)Be than previously reported soils. (10)Be may be useful for measuring the fast neutron dose near the Hiroshima bomb hypocenter at neutron energies double those previously available.

Porous carbon-coated graphite electrodes for energy production from salinity gradient using reverse electrodialysis

NASA Astrophysics Data System (ADS)

Lee, Su-Yoon; Jeong, Ye-Jin; Chae, So-Ryong; Yeon, Kyeong-Ho; Lee, Yunkyu; Kim, Chan-Soo; Jeong, Nam-Jo; Park, Jin-Soo

2016-04-01

Performance of graphite foil electrodes coated by porous carbon black (i.e., Vulcan) was investigated in comparison with metal electrodes for reverse electrodialysis (RED) application. The electrode slurry that was used for fabrication of the porous carbon-coated graphite foil is composed of 7.2 wt% of carbon black (Vulcan X-72), 0.8 wt% of a polymer binder (polyvinylidene fluoride, PVdF), and 92.0 wt% of a mixing solvent (dimethylacetamide, DMAc). Cyclic voltammograms of both the porous carbon (i.e., Vulcan)-coated graphite foil electrode and the graphite foil electrode without Vulcan showed good reversibility in the hexacyanoferrate(III) (i.e., Fe(CN)63-) and hexacyanoferrate(II) (i.e., Fe(CN)64-) redox couple and 1 M Na2SO4 at room temperature. However, anodic and cathodic current of the Vulcan-coated graphite foil electrode was much higher than those of the graphite foil electrode. Using a bench-scale RED stack, the current-voltage polarization curve of the Vulcan-coated graphite electrode was compared to that of metal electrodes such as iridium (Ir) and platinum (Pt). From the results, it was confirmed that resistance of four different electrodes increased with the following order: the Vulcan-coated graphite foil

Influence of graphite-alloy interactions on corrosion of Ni-Mo-Cr alloy in molten fluorides

NASA Astrophysics Data System (ADS)

Ai, Hua; Hou, Juan; Ye, Xiang-Xi; Zeng, Chao Liu; Sun, Hua; Li, Xiaoyun; Yu, Guojun; Zhou, Xingtai; Wang, Jian-Qiang

2018-05-01

In this study, the effects of graphite-alloy interaction on corrosion of Ni-Mo-Cr alloy in molten FLiNaK salt were investigated. The corrosion tests of Ni-Mo-Cr alloys were conducted in graphite crucibles, to examine the differences of test specimens in conditions of electric contact and isolated with graphite, respectively. The corrosion attack is severer with more weight loss and deeper Cr depletion layer in samples electric contact with graphite than those isolated with graphite. The occurrence of galvanic corrosion between alloy specimens and graphite container was confirmed by electrochemical measurement. The corrosion is controlled by nonelectric transfer in isolated test while electrochemical reaction accelerated corrosion in electric contact test.

The origin of epigenetic graphite: evidence from isotopes

USGS Publications Warehouse

Weis, P.L.; Friedman, I.; Gleason, J.P.

1981-01-01

Stable carbon isotope ratios measured in syngenetic graphite, epigenetic graphite, and graphitic marble suggests that syngenetic graphite forms only by the metamorphism of carbonaceous detritus. Metamorphism of calcareous rocks with carbonaceous detritus is accompanied by an exchange of carbon between the two, which may result in large changes in isotopic composition of the non-carbonate phase but does not affect the relative proportions of the two reactants in the rock. Epigenetic graphite forms only from carbonaceous material or preexisting graphite. The reactions involved are the water gas reaction (C + H2O ??? CO + H2) at 800-900??C, and the Boudouard reaction (2CO ??? C + CO2), which probably takes place at temperatures about 50-100??C lower. ?? 1982.

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

DOEpatents

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.

METHOD FOR COATING GRAPHITE WITH METALLIC CARBIDES

DOEpatents

Steinberg, M.A.

1960-03-22

A method for producing refractory coatings of metallic carbides on graphite was developed. In particular, the graphite piece to be coated is immersed in a molten solution of 4 to 5% by weight of zirconium, titanium, or niobium dissolved in tin. The solution is heated in an argon atmosphere to above 1400 deg C, whereby the refractory metal reacts with the surface of the graphite to form a layer of metalic carbide. The molten solution is cooled to 300 to 400 deg C, and the graphite piece is removed. Excess tin is wiped from the graphite, which is then heated in vacuum to above 2300 deg C. The tin vaporizes from the graphite surface, leaving the surface coated with a tenacious layer of refractory metallic carbide.

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

PubMed

Srivastava, Suneel Kumar; Pionteck, Jürgen

2015-03-01

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

The Sustainable Nuclear Future: Fission and Fusion E.M. Campbell Logos Technologies

NASA Astrophysics Data System (ADS)

Campbell, E. Michael

2010-02-01

Global industrialization, the concern over rising CO2 levels in the atmosphere and other negative environmental effects due to the burning of hydrocarbon fuels and the need to insulate the cost of energy from fuel price volatility have led to a renewed interest in nuclear power. Many of the plants under construction are similar to the existing light water reactors but incorporate modern engineering and enhanced safety features. These reactors, while mature, safe and reliable sources of electrical power have limited efficiency in converting fission power to useful work, require significant amounts of water, and must deal with the issues of nuclear waste (spent fuel), safety, and weapons proliferation. If nuclear power is to sustain its present share of the world's growing energy needs let alone displace carbon based fuels, more than 1000 reactors will be needed by mid century. For this to occur new reactors that are more efficient, versatile in their energy markets, require minimal or no water, produce less waste and more robust waste forms, are inherently safe and minimize proliferation concerns will be necessary. Graphite moderated, ceramic coated fuel, and He cooled designs are reactors that can satisfy these requirements. Along with other generation IV fast reactors that can further reduce the amounts of spent fuel and extend fuel resources, such a nuclear expansion is possible. Furthermore, facilities either in early operations or under construction should demonstrate the next step in fusion energy development in which energy gain is produced. This demonstration will catalyze fusion energy development and lead to the ultimate development of the next generation of nuclear reactors. In this presentation the role of advanced fission reactors and future fusion reactors in the expansion of nuclear power will be discussed including synergies with the existing worldwide nuclear fleet. )

CMB-13 research on carbon and graphite

NASA Technical Reports Server (NTRS)

Smith, M. C.

1972-01-01

The research on graphite and carbon for this period is reported. Topics discussed include: effects of grinding on the Santa Marie graphites, properties and purities of coal-tar, resin-bonded graphite, carbonization of resin components, and glass-like carbon filler.

Tribological Analysis of Copper-Coated Graphite Particle-Reinforced A359 Al/5 wt.% SiC Composites

NASA Astrophysics Data System (ADS)

Lin, C. B.; Wang, T. C.; Chang, Z. C.; Chu, H. Y.

2013-01-01

Copper-coated graphite particles can be mass-produced by the cementation process using simple equipment. Graphite particulates that were coated with electroless copper and 5 wt.% SiC particulates were introduced into an aluminum alloy by compocasting to make A359 Al/5 wt.% SiC(p) composite that contained 2, 4, 6, and 8 wt.% graphite particulate composite. The effects of SiC particles, quantity of graphite particles, normal loading, sliding speed and wear debris on the coefficient of friction, and the wear rate were investigated. The results thus obtained indicate that the wear properties were improved by adding small amounts of SiC and graphite particles into the A359 Al alloy. The coefficient of friction of the A359 Al/5 wt.% SiC(p) composite that contained 6.0 wt.% graphite particulates was reduced to 0.246 and the amount of graphite film that was released on the worn surface increased with the graphite particulate content. The coefficient of friction and the wear rate were insensitive to the variation in the sliding speed and normal loading.

Preparation and characterization of copper-graphite composites by electrical explosion of wire in liquid.

PubMed

Bien, T N; Gul, W H; Bac, L H; Kim, J C

2014-11-01

Copper-graphite nanocomposites containing 5 vol.% graphite were prepared by a powder metallurgy route using an electrical wire explosion (EEW) in liquid method and spark plasma sintering (SPS) process. Graphite rods with a 0.3 mm diameter and copper wire with a 0.2 mm diameter were used as raw materials for EEWin liquid. To compare, a pure copper and copper-graphite mixture was also prepared. The fabricated graphite was in the form of a nanosheet, onto which copper particles were coated. Sintering was performed at 900 degrees C at a heating rate of 30 degrees C/min for 10 min and under a pressure of 70 MPa. The density of the sintered composite samples was measured by the Archimedes method. A wear test was performed by a ball-on-disc tribometer under dry conditions at room temperature in air. The presence of graphite effectively reduced the wear of composites. The copper-graphite nanocomposites prepared by EEW had lower wear rates than pure copper material and simple mixed copper-graphite.

Understanding the crack formation of graphite particles in cycled commercial lithium-ion batteries by focused ion beam - scanning electron microscopy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lin, Na; Jia, Zhe; Wang, Zhihui

Here in this paper, the structure degradation of commercial Lithium-ion battery (LIB) graphite anodes with different cycling numbers and charge rates was investigated by focused ion beam (FIB) and scanning electron microscopy (SEM). The cross-section image of graphite anode by FIB milling shows that cracks, resulted in the volume expansion of graphite electrode during long-term cycling, were formed in parallel with the current collector. The crack occurs in the bulk of graphite particles near the lithium insertion surface, which might derive from the stress induced during lithiation and de-lithiation cycles. Subsequently, crack takes place along grain boundaries of the polycrystallinemore » graphite, but only in the direction parallel with the current collector. Furthermore, fast charge graphite electrodes are more prone to form cracks since the tensile strength of graphite is more likely to be surpassed at higher charge rates. Therefore, for LIBs long-term or high charge rate applications, the tensile strength of graphite anode should be taken into account.« less

Understanding the crack formation of graphite particles in cycled commercial lithium-ion batteries by focused ion beam - scanning electron microscopy

DOE PAGES

Lin, Na; Jia, Zhe; Wang, Zhihui; ...

2017-10-01

Here in this paper, the structure degradation of commercial Lithium-ion battery (LIB) graphite anodes with different cycling numbers and charge rates was investigated by focused ion beam (FIB) and scanning electron microscopy (SEM). The cross-section image of graphite anode by FIB milling shows that cracks, resulted in the volume expansion of graphite electrode during long-term cycling, were formed in parallel with the current collector. The crack occurs in the bulk of graphite particles near the lithium insertion surface, which might derive from the stress induced during lithiation and de-lithiation cycles. Subsequently, crack takes place along grain boundaries of the polycrystallinemore » graphite, but only in the direction parallel with the current collector. Furthermore, fast charge graphite electrodes are more prone to form cracks since the tensile strength of graphite is more likely to be surpassed at higher charge rates. Therefore, for LIBs long-term or high charge rate applications, the tensile strength of graphite anode should be taken into account.« less

Hybridized polymer matrix composites

NASA Technical Reports Server (NTRS)

House, E. E.; Hoggatt, J. T.; Symonds, W. A.

1980-01-01

The extent to which graphite fibers are released from resin matrix composites that are exposed to fire and impact conditions was determined. Laboratory simulations of those conditions that could exist in the event of an aircraft crash and burn situation were evaluated. The effectiveness of various hybridizing concepts in preventing this release of graphite fibers were also evaluated. The baseline (i.e., unhybridized) laminates examined were prepared from commercially available graphite/epoxy, graphite/polyimide, and graphite/phenolic materials. Hybridizing concepts investigated included resin fillers, laminate coatings, resin blending, and mechanical interlocking of the graphite reinforcement. The baseline and hybridized laminates' mechanical properties, before and after isothermal and humidity aging, were also compared. It was found that a small amount of graphite fiber was released from the graphite/epoxy laminates during the burn and impact conditions used in this program. However, the extent to which the fibers were released is not considered a severe enough problem to preclude the use of graphite reinforced composites in civil aircraft structure. It also was found that several hybrid concepts eliminated this fiber release. Isothermal and humidity aging did not appear to alter the fiber release tendencies.

Self-Propagating Combustion Triggered Synthesis of 3D Lamellar Graphene/BaFe12O19 Composite and Its Electromagnetic Wave Absorption Properties

PubMed Central

Zhao, Tingkai; Ji, Xianglin; Jin, Wenbo; Yang, Wenbo; Peng, Xiarong; Duan, Shichang; Dang, Alei; Li, Hao; Li, Tiehu

2017-01-01

The synthesis of 3D lamellar graphene/BaFe12O19 composites was performed by oxidizing graphite and sequentially self-propagating combustion triggered process. The 3D lamellar graphene structures were formed due to the synergistic effect of the tremendous heat induced gasification as well as huge volume expansion. The 3D lamellar graphene/BaFe12O19 composites bearing 30 wt % graphene present the reflection loss peak at −27.23 dB as well as the frequency bandwidth at 2.28 GHz (< −10 dB). The 3D lamellar graphene structures could consume the incident waves through multiple reflection and scattering within the layered structures, prolonging the propagation path of electromagnetic waves in the absorbers. PMID:28336889

High speed hydrogen/graphite interaction

NASA Technical Reports Server (NTRS)

Kelly, A. J.; Hamman, R.; Sharma, O. P.; Harrje, D. T.

1974-01-01

Various aspects of a research program on high speed hydrogen/graphite interaction are presented. Major areas discussed are: (1) theoretical predictions of hydrogen/graphite erosion rates; (2) high temperature, nonequilibrium hydrogen flow in a nozzle; and (3) molecular beam studies of hydrogen/graphite erosion.

RECOVERY OF VALUABLE MATERIAL FROM GRAPHITE BODIES

DOEpatents

Fromm, L.W. Jr.

1959-09-01

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

Interface Character of Aluminum-Graphite Metal Matrix Composites.

DTIC Science & Technology

1983-01-27

studied included the commer- cial A/graphite composites; layered model systems on single crystal and poly- crystalline graphite substrates as well as...composition and thickness of the composite interface, and graphite crystal orientation. 3 For the model systems in this study , single crystal graphite...been reviewed by Kingcry. Segregation at surfaces in single- crystal MgO of Fe, Cr and Sc, which were Dresent in concentrations within the single- 3phase

Structure and Performance of Epoxy Resin Cladded Graphite Used as Anode

NASA Astrophysics Data System (ADS)

Zhou, Zhentao; Li, Haijun

This paper is concerning to prepare modified natural graphite which is low-cost and advanced materials used as lithium ion battery anode using the way of cladding natural graphite with epoxy resin. The results shows that the specific capacity and circular performance of the modified natural graphite, which is prepared in the range of 600°C and 1000°C, have been apparently improved compare with the not-modified natural graphite. The first reversible capacity of the modified natural graphite is 338mAh/g and maintain more than 330mAh/g after 20 charge/discharge circles.

Artificially-built solid electrolyte interphase via surface-bonded vinylene carbonate derivative on graphite by molecular layer deposition

NASA Astrophysics Data System (ADS)

Chae, Seulki; Lee, Jeong Beom; Lee, Jae Gil; Lee, Tae-jin; Soon, Jiyong; Ryu, Ji Heon; Lee, Jin Seok; Oh, Seung M.

2017-12-01

Vinylene carbonate (VC) is attached in a ring-opened form on a graphite surface by molecular layer deposition (MLD) method, and its role as a solid electrolyte interphase (SEI) former is studied. When VC is added into the electrolyte solution of a graphite/LiNi0.5Mn1.5O4 (LNMO) full-cell, it is reductively decomposed to form an effective SEI on the graphite electrode. However, VC in the electrolyte solution has serious adverse effects due to its poor stability against electrochemical oxidation on the LNMO positive electrode. A excessive acid generation as a result of VC oxidation is observed, causing metal dissolution from the LNMO electrode. The dissolved metal ions are plated on the graphite electrode to destroy the SEI layer, eventually causing serious capacity fading and poor Coulombic efficiency. The VC derivative on the graphite surface also forms an effective SEI layer on the graphite negative electrode via reductive decomposition. The detrimental effects on the LNMO positive electrode, however, can be avoided because the bonded VC derivative on the graphite surface cannot move to the LNMO electrode. Consequently, the graphite/LNMO full-cell fabricated with the VC-attached graphite outperforms the cells without VC or with VC in the electrolyte, in terms of Coulombic efficiency and capacity retention.

Micro-fabrication method of graphite mesa microdevices based on optical lithography technology

NASA Astrophysics Data System (ADS)

Zhang, Cheng; Wen, Donghui; Zhu, Huamin; Zhang, Xiaorui; Yang, Xing; Shi, Yunsheng; Zheng, Tianxiang

2017-12-01

Graphite mesa microdevices have incommensurate contact nanometer interfaces, superlubricity, high-speed self-retraction, and other characteristics, which have potential applications in high-performance oscillators and micro-scale switches, memory devices, and gyroscopes. However, the current method of fabricating graphite mesa microdevices is mainly based on high-cost, low efficiency electron beam lithography technology. In this paper, the processing technologies of graphite mesa microdevices with various shapes and sizes were investigated by a low-cost micro-fabrication method, which was mainly based on optical lithography technology. The characterization results showed that the optical lithography technology could realize a large-area of patterning on the graphite surface, and the graphite mesa microdevices, which have a regular shape, neat arrangement, and high verticality could be fabricated in large batches through optical lithography technology. The experiments and analyses showed that the graphite mesa microdevices fabricated through optical lithography technology basically have the same self-retracting characteristics as those fabricated through electron beam lithography technology, and the maximum size of the graphite mesa microdevices with self-retracting phenomenon can reach 10 µm  ×  10 µm. Therefore, the proposed method of this paper can realize the high-efficiency and low-cost processing of graphite mesa microdevices, which is significant for batch fabrication and application of graphite mesa microdevices.

Graphite tail powder and liquid biofertilizer as trace elements source for ground nut

NASA Astrophysics Data System (ADS)

Hindersah, Reginawanti; Setiawati, M. Rochimi; Fitriatin, B. Natalie; Suryatama, Pujawati; Asmiran, Priyanka; Panatarani, Camellia; Joni, I. Made

2018-02-01

Utilization of graphite tail waste from the mineral beneficiation processing is very important since it contain significant amount of essential minerals which are necessary for plant growth. These mineral are required in biochemical processes and mainly play an important role as cofactor in enzymatic reaction. The objective of this research is to investigate the performance of graphite tail on supporting plant growth and yield of ground nut (Arachishypogeae L.). A field experiment has been performed to test the performance of mixed graphite tail and reduced organic matter dose. The graphite tail size were reduced to various sieved size, -80 mesh, -100 mesh and -200 mesh. The experiment was setup in randomized block design with 4 treatments and 6 replications for each treatment, while the control plot is received without graphite tail. The results demonstrated that reduced organic matter along with -200 mesh tail has potentially decreased plant height at the end of vegetative growth stage, in contrast for to -80 mesh tail amendment increased individual fresh plant biomass. Statistically, there was no change of plant nodule, individual shoot fresh and dry weight, root nodule, number of pod following any mesh of graphite tail amendment. Reducing organic matter while adding graphite tail of 5% did not change bean weight in all plot. In contrast, reduced organic matter along with 80-mesh graphite tail amendment improved the nut yield per plot. This experiment suggests that graphite tail, mainly -80 mesh graphite tail can be possibly used in legume production.

40 CFR 436.380 - Applicability; description of the graphite subcategory.

Code of Federal Regulations, 2011 CFR

2011-07-01

... graphite subcategory. 436.380 Section 436.380 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) EFFLUENT GUIDELINES AND STANDARDS MINERAL MINING AND PROCESSING POINT SOURCE CATEGORY Graphite Subcategory § 436.380 Applicability; description of the graphite subcategory. The provisions of this subpart...

40 CFR 436.380 - Applicability; description of the graphite subcategory.

Code of Federal Regulations, 2010 CFR

2010-07-01

... graphite subcategory. 436.380 Section 436.380 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) EFFLUENT GUIDELINES AND STANDARDS MINERAL MINING AND PROCESSING POINT SOURCE CATEGORY Graphite Subcategory § 436.380 Applicability; description of the graphite subcategory. The provisions of this subpart...

Temperature effect of friction and wear characteristics for solid lubricating graphite

NASA Astrophysics Data System (ADS)

Kim, Yeonwook; Kim, Jaehoon

2015-03-01

Graphite is one of the effective lubricant additives due to its excellent high-temperature endurance and self-lubricating properties. In this study, wear behavior of graphite used as sealing materials to cut off hot gas is evaluated at room and elevated temperature. Wear occurs on graphite seal due to the friction of driving shaft and graphite. Thus, a reciprocating wear test to evaluate the wear generated for the graphite by means of the relative motion between a shaft material and a graphite seal was carried out. The friction coefficient and specific wear rate for the changes of applied load and sliding speed were compared under different temperature conditions considering the actual operating environment. Through SEM observation of the worn surface, the lubricating film was observed and compared with test conditions.

Fabrication and testing of non-graphitic superhybrid composites

NASA Technical Reports Server (NTRS)

Lark, R. F.; Sinclair, J. H.; Chamis, C. C.

1979-01-01

A study was conducted to determine the fabrication feasibility and the mechanical properties of adhesively-bonded boron aluminum/titanium and non-graphitic fiber/epoxy resin superhybrid (NGSH) composite laminates for potential aerospace applications. The major driver for this study was the elimination of a potential graphite fiber release problem in the event of a fire. The results of the study show that non-graphitic fibers, such as S-glass and Kevlar 49, may be substituted for the graphite fibers used in superhybrid (SH) composites for some applications. As is to be expected, however, the non-graphitic superhybrids have lower stiffness properties than the graphitic superhybrids. In-plane and flexural moduli of the laminates studied in this program can be predicted reasonably well using linear laminate theory while nonlinear laminate theory is required for strength predictions.

Absolute dosimetry on a dynamically scanned sample for synchrotron radiotherapy using graphite calorimetry and ionization chambers

NASA Astrophysics Data System (ADS)

Lye, J. E.; Harty, P. D.; Butler, D. J.; Crosbie, J. C.; Livingstone, J.; Poole, C. M.; Ramanathan, G.; Wright, T.; Stevenson, A. W.

2016-06-01

The absolute dose delivered to a dynamically scanned sample in the Imaging and Medical Beamline (IMBL) on the Australian Synchrotron was measured with a graphite calorimeter anticipated to be established as a primary standard for synchrotron dosimetry. The calorimetry was compared to measurements using a free-air chamber (FAC), a PTW 31 014 Pinpoint ionization chamber, and a PTW 34 001 Roos ionization chamber. The IMBL beam height is limited to approximately 2 mm. To produce clinically useful beams of a few centimetres the beam must be scanned in the vertical direction. In practice it is the patient/detector that is scanned and the scanning velocity defines the dose that is delivered. The calorimeter, FAC, and Roos chamber measure the dose area product which is then converted to central axis dose with the scanned beam area derived from Monte Carlo (MC) simulations and film measurements. The Pinpoint chamber measures the central axis dose directly and does not require beam area measurements. The calorimeter and FAC measure dose from first principles. The calorimetry requires conversion of the measured absorbed dose to graphite to absorbed dose to water using MC calculations with the EGSnrc code. Air kerma measurements from the free air chamber were converted to absorbed dose to water using the AAPM TG-61 protocol. The two ionization chambers are secondary standards requiring calibration with kilovoltage x-ray tubes. The Roos and Pinpoint chambers were calibrated against the Australian primary standard for air kerma at the Australian Radiation Protection and Nuclear Safety Agency (ARPANSA). Agreement of order 2% or better was obtained between the calorimetry and ionization chambers. The FAC measured a dose 3-5% higher than the calorimetry, within the stated uncertainties.

A contrastive study of three graphite anodes in the piperidinium based electrolytes for lithium ion batteries

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiang, Xiao-Tao; Wang, Chen-Yi; Gao, Kun, E-mail: gaokun0451@163.com

Graphical abstract: The fitting results of R{sub sei} and R{sub ct} of three graphite/Li cells. Besides three graphite/Li cells show the similar R{sub sei}, the NG198/Li cell demonstrates a higher R{sub ct} value in all test temperatures. Especially, the R{sub ct} at 333 K is even up to 355.8 Ω cm{sup 2}. Obviously, the narrow distribution of edge plane for NG198 caused this result, and then greatly restricts its cell capacity. By contrast, CMB with bigger specific surface area and more Li{sup +} insertion points shows lower resistance at room temperature, which should help to improve its capacity. - Highlights:more » • SEI film is closely related to graphite structures and formation temperature. • The graphite with bigger surface area and more Li{sup +} insertion points behaves better. • The graphite with narrow edge plane is uncompetitive for ionic liquid electrolyte. - Abstract: The electrochemical behaviors of natural graphite (NG198), artificial graphite (AG360) and carbon microbeads (CMB) in an ionic liquid based electrolyte are investigated by cyclic voltammetry (CV). The surface and structure of three graphite materials are characterized by scanning electron microscope (SEM) and X-ray diffraction (XRD) before and after cycling. It is found that solid electrolyte interface (SEI) is closely related to graphite structure. Benefiting from larger specific surface area and more dispersed Li{sup +} insertion points, CMB shows a better Li{sup +} insertion/de-insertion behavior than NG198 and AG360. Furthermore, electrochemical impedance spectra (EIS) prove that the SEI of different graphite electrodes has different intrinsic resistance and Li{sup +} penetrability. By comparison, CMB behaves better cell performances than AG360, while the narrow edge plane makes NG198 uncompetitive as a potential anode for the ionic liquids (ILs)-type Li-ion battery.« less

Is Water at the Graphite Interface Vapor-like or Ice-like?

PubMed

Qiu, Yuqing; Lupi, Laura; Molinero, Valeria

2018-04-05

Graphitic surfaces are the main component of soot, a major constituent of atmospheric aerosols. Experiments indicate that soots of different origins display a wide range of abilities to heterogeneously nucleate ice. The ability of pure graphite to nucleate ice in experiments, however, seems to be almost negligible. Nevertheless, molecular simulations with the monatomic water model mW with water-carbon interactions parameterized to reproduce the experimental contact angle of water on graphite predict that pure graphite nucleates ice. According to classical nucleation theory, the ability of a surface to nucleate ice is controlled by the binding free energy between ice immersed in liquid water and the surface. To establish whether the discrepancy in freezing efficiencies of graphite in mW simulations and experiments arises from the coarse resolution of the model or can be fixed by reparameterization, it is important to elucidate the contributions of the water-graphite, water-ice, and ice-water interfaces to the free energy, enthalpy, and entropy of binding for both water and the model. Here we use thermodynamic analysis and free energy calculations to determine these interfacial properties. We demonstrate that liquid water at the graphite interface is not ice-like or vapor-like: it has similar free energy, entropy, and enthalpy as water in the bulk. The thermodynamics of the water-graphite interface is well reproduced by the mW model. We find that the entropy of binding between graphite and ice is positive and dominated, in both experiments and simulations, by the favorable entropy of reducing the ice-water interface. Our analysis indicates that the discrepancy in freezing efficiencies of graphite in experiments and the simulations with mW arises from the inability of the model to simultaneously reproduce the contact angle of liquid water on graphite and the free energy of the ice-graphite interface. This transferability issue is intrinsic to the resolution of the model, and arises from its lack of rotational degrees of freedom.

Dirty snow after nuclear war

NASA Technical Reports Server (NTRS)

Warren, S. G.; Wiscombe, W. J.

1985-01-01

It is shown that smoke from fires started by nuclear explosions could continue to cause significant disruption even after it has fallen from the atmosphere, by lowering the reflectivity of snow and sea ice surfaces, with possible effects on climate in northern latitudes caused by enhanced absorption of sunlight. The reduced reflectivity could persist for several years on Arctic sea ice and on the ablation area of the Greenland ice sheet.

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

PubMed

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.

Direct Preparation of Few Layer Graphene Epoxy Nanocomposites from Untreated Flake Graphite.

PubMed

Throckmorton, James; Palmese, Giuseppe

2015-07-15

The natural availability of flake graphite and the exceptional properties of graphene and graphene-polymer composites create a demand for simple, cost-effective, and scalable methods for top-down graphite exfoliation. This work presents a novel method of few layer graphite nanocomposite preparation directly from untreated flake graphite using a room temperature ionic liquid and laminar shear processing regimen. The ionic liquid serves both as a solvent and initiator for epoxy polymerization and is incorporated chemically into the matrix. This nanocomposite shows low electrical percolation (0.005 v/v) and low thickness (1-3 layers) graphite/graphene flakes by TEM. Additionally, the effect of processing conditions by rheometry and comparison with solvent-free conditions reveal the interactions between processing and matrix properties and provide insight into the theory of the chemical and physical exfoliation of graphite crystals and the resulting polymer matrix dispersion. An interaction model that correlates the interlayer shear physics of graphite flakes and processing parameters is proposed and tested.

Adsorption behavior of bisphenol A on CTAB-modified graphite

NASA Astrophysics Data System (ADS)

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.

Phase Structures and Magnetic Properties of Graphite Nanosheets and Ni-Graphite Nanocomposite Synthesized by Electrical Explosion of Wire in Liquid

NASA Astrophysics Data System (ADS)

Nguyen, Minh-Thuyet; Kim, Jin-Hyung; Lee, Jung-Goo; Kim, Jin-Chun

2018-03-01

The present work studied on phases and magnetic properties of graphite nanosheets and Ni-graphite nanocomposite synthesized using the electrical explosion of wire (EEW) in ethanol. X-ray diffraction and field emission scanning electron microscope were used to investigate the phases and the morphology of the nanopowders obtained. It was found that graphite nanosheets were absolutely fabricated by EEW with a thickness of 29 nm and 3 μm diameter. The as-synthesized Ni-graphite composite powders had a Ni-coating on the surfaces of graphite sheets. The hysteresis loop of the as-exploded, the hydrogen-treated composite nanopowders and the sintered samples were examined with a vibrating sample magnetometer at room temperature. The Ni-graphite composite exposed the magnetic behaviors which are attributed to Ni component. The magnetic properties of composite had the improvement from 10.2 emu/g for the as-exploded powders to 15.8 emu/g for heat-treated powders and 49.16 emu/g for sintered samples.

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

NASA Astrophysics Data System (ADS)

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.

Fire test method for graphite fiber reinforced plastics

NASA Technical Reports Server (NTRS)

Bowles, K. J.

1980-01-01

A potential problem in the use of graphite fiber reinforced resin matrix composites is the dispersal of graphite fibers during accidential fires. Airborne, electrically conductive fibers originating from the burning composites could enter and cause shorting in electrical equipment located in surrounding areas. A test method for assessing the burning characteristics of graphite fiber reinforced composites and the effectiveness of the composites in retaining the graphite fibers has been developed. The method utilizes a modified rate of heat release apparatus. The equipment and the testing procedure are described. The application of the test method to the assessment of composite materials is illustrated for two resin matrix/graphite composite systems.

Process for the fabrication of aluminum metallized pyrolytic graphite sputtering targets

DOEpatents

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

1995-01-01

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

Arsenic Removal from Water by Adsorption on Iron-Contaminated Cryptocrystalline Graphite

NASA Astrophysics Data System (ADS)

Yang, Qiang; Yang, Lang; Song, Shaoxian; Xia, Ling

This work aimed to study the feasibility of using iron-contaminated graphite as an adsorbent for As(V) removal from water. The adsorbent was prepared by grinding graphite concentrate with steel ball. The study was performed through the measurements of adsorption capacity, BET surface area and XPS analysis. The experimental results showed that the iron-contaminated graphite exhibited significantly high adsorption capacity of As(V). The higher the iron contaminated on the graphite surface, the higher the adsorption capacity of As(V) on the material obtained. It was suggested that the ion-contaminated graphite was a good adsorbent for As(V) removal.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Inam, A., E-mail: aqil.ceet@pu.edu.pk; Brydson, R., E-mail: mtlrmdb@leeds.ac.uk; Edmonds, D.V., E-mail: d.v.edmonds@leeds.ac.uk

The potential for using graphite particles as an internal lubricant during machining is considered. Graphite particles were found to form during graphitisation of experimental medium-carbon steel alloyed with Si and Al. The graphite nucleation sites were strongly influenced by the starting microstructure, whether ferrite–pearlite, bainite or martensite, as revealed by light and electron microscopy. Favourable nucleation sites in the ferrite–pearlite starting microstructure were, not unexpectedly, found to be located within pearlite colonies, no doubt due to the presence of abundant cementite as a source of carbon. In consequence, the final distribution of graphite nodules in ferrite–pearlite microstructures was less uniformmore » than for the bainite microstructure studied. In the case of martensite, this study found a predominance of nucleation at grain boundaries, again leading to less uniform graphite dispersions. - Highlights: • Metallography of formation of graphite particles in experimental carbon steel. • Potential for using graphite in steel as an internal lubricant during machining. • Microstructure features expected to influence improved machinability studied. • Influence of pre-anneal starting microstructure on graphite nucleation sites. • Influence of pre-anneal starting microstructure on graphite distribution. • Potential benefit is new free-cutting steel compositions without e.g. Pb alloying.« less

S&TR Preview: Smashing Materials to Reveal Unusual Behavior

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunsberger, Maren; Akin, Minta; Chau, Ricky

2016-01-13

Squeeze a material hard enough, and its structure and properties will change, sometimes dramatically so. With enough heat and pressure, scientists can turn pencil lead (graphite), one of Earth’s softest materials, into diamond, one of its hardest. Apply even more pressure—such as might be found in explosions, detonating nuclear weapons, laser fusion experiments, meteorite impacts, or the hearts of stars and planets—and materials can take stranger forms. Deep in Jupiter’s core, for instance, where pressures likely reach 50 to 100 million times that of Earth’s atmosphere, hydrogen is predicted to be a metallic liquid rather than the familiar transparent gas.

Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

NASA Astrophysics Data System (ADS)

Krása, Antonín; Kochetkov, Anatoly; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente

2017-09-01

VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector). Discrepancies between experiments and Monte Carlo calculations (MCNP5) of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler) depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2) are presented.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chuang, Chihpin; Singh, Dileep; Kenesei, Peter

The size and morphology of the graphite particles play a crucial role in determining various mechanical and thermal properties of cast iron. In the present study, we utilized high-energy synchrotron X-ray tomography to perform quantitative 3D-characterization of the distribution of graphite particles in high-strength compacted graphite iron (CGI). The size, shape, and spatial connectivity of graphite were examined. The analysis reveals that the compacted graphite can grow with a coral-tree-like morphology and span several hundred microns in the iron matrix.

Preparation of pyrolytic carbon coating on graphite for inhibiting liquid fluoride salt and Xe135 penetration for molten salt breeder reactor

NASA Astrophysics Data System (ADS)

Song, Jinliang; Zhao, Yanling; He, Xiujie; Zhang, Baoliang; Xu, Li; He, Zhoutong; Zhang, DongSheng; Gao, Lina; Xia, Huihao; Zhou, Xingtai; Huai, Ping; Bai, Shuo

2015-01-01

A fixed-bed deposition method was used to prepare rough laminar pyrolytic carbon coating (RLPyC) on graphite for inhibiting liquid fluoride salt and Xe135 penetration during use in molten salt breeder reactor. The RLPyC coating possessed a graphitization degree of 44% and had good contact with graphite substrate. A high-pressure reactor was constructed to evaluate the molten salt infiltration in the isostatic graphite (IG-110, TOYO TANSO CO., LTD.) and RLPyC coated graphite under 1.01, 1.52, 3.04, 5.07 and 10.13 × 105 Pa for 12 h. Mercury injection and molten-salt infiltration experiments indicated the porosity and the salt-infiltration amount of 18.4% and 13.5 wt% under 1.52 × 105 Pa of IG-110, which was much less than 1.2% and 0.06 wt% under 10.13 × 105 Pa of the RLPyC, respectively. A vacuum device was constructed to evaluate the Xe135 penetration in the graphite. The helium diffusion coefficient of RLPyC coated graphite was 2.16 × 10-12 m2/s, much less than 1.21 × 10-6 m2/s of the graphite. Thermal cycle experiment indicated the coatings possessed excellent thermal stability. The coated graphite could effectively inhibit the liquid fluoride salt and Xe135 penetration.

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

NASA Astrophysics Data System (ADS)

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

2018-06-01

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

Graphite fiber reinforced thermoplastic resins

NASA Technical Reports Server (NTRS)

Navak, R. C.

1977-01-01

The results of a program designed to optimize the fabrication procedures for graphite thermoplastic composites are described. The properties of the composites as a function of temperature were measured and graphite thermoplastic fan exit guide vanes were fabricated and tested. Three thermoplastics were included in the investigation: polysulfone, polyethersulfone, and polyarylsulfone. Type HMS graphite was used as the reinforcement. Bending fatigue tests of HMS graphite/polyethersulfone demonstrated a gradual shear failure mode which resulted in a loss of stiffness in the specimens. Preliminary curves were generated to show the loss in stiffness as a function of stress and number of cycles. Fan exit guide vanes of HMS graphite polyethersulfone were satisfactorily fabricated in the final phase of the program. These were found to have stiffness and better fatigue behavior than graphite epoxy vanes which were formerly bill of material.

High efficiency of CO2-activated graphite felt as electrode for vanadium redox flow battery application

NASA Astrophysics Data System (ADS)

Chang, Yu-Chung; Chen, Jian-Yu; Kabtamu, Daniel Manaye; Lin, Guan-Yi; Hsu, Ning-Yih; Chou, Yi-Sin; Wei, Hwa-Jou; Wang, Chen-Hao

2017-10-01

A simple method for preparing CO2-activated graphite felt as an electrode in a vanadium redox flow battery (VRFB) was employed by the direct treatment in a CO2 atmosphere at a high temperature for a short period. The CO2-activated graphite felt demonstrates excellent electrochemical activity and reversibility. The VRFB using the CO2-activated graphite felts in the electrodes has coulombic, voltage, and energy efficiencies of 94.52%, 88.97%, and 84.15%, respectively, which is much higher than VRFBs using the electrodes of untreated graphite felt and N2-activated graphite felt. The efficiency enhancement was attributed to the higher number of oxygen-containing functional groups on the graphite felt that are formed during the CO2-activation, leading to improving the electrochemical behaviour of the resultant VRFB.

Influence of the Interaction Between Graphite and Polar Surfaces of ZnO on the Formation of Schottky Contact

NASA Astrophysics Data System (ADS)

Yatskiv, R.; Grym, J.

2018-03-01

We show that the interaction between graphite and polar surfaces of ZnO affects electrical properties of graphite/ZnO Schottky junctions. A strong interaction of the Zn-face with the graphite contact causes interface imperfections and results in the formation of laterally inhomogeneous Schottky contacts. On the contrary, high quality Schottky junctions form on the O-face, where the interaction is significantly weaker. Charge transport through the O-face ZnO/graphite junctions is well described by the thermionic emission model in both forward and reverse directions. We further demonstrate that the parameters of the graphite/ZnO Schottky diodes can be significantly improved when a thin layer of ZnO2 forms at the interface between graphite and ZnO after hydrogen peroxide surface treatment.

Thermophysical properties of graphite HOPG and HAPG in the solid state and under melting (from 2000 K up to 5000 K)

NASA Astrophysics Data System (ADS)

Savvatimskiy, A. I.; Onufriev, S. V.; Konyukhov, S. A.

2017-11-01

Experiments with HOPG graphite grade showed that the melting temperature of graphite equals 4800-4900 K and that the melting of graphite is possible only at elevated pressures. The data were obtained for resistivity, specific heat and input (Joule) energy up to 5000 K. HAPG (Highly Annealing Pyrolytic Graphite) is a form of highly oriented pyrolytic graphite. HAPG specimens in the form of strips (thickness 30 microns) were placed in a cell (between two plates of glass-sapphire). The specimen temperature was measured by a high speed pyrometer. The heat of fusion for both graphite grades (heated in a confined volume) was less (and specific heat - higher) than for the case with nearly free expansion. A possible reason for the observed effects is discussed in the report.

Modelling deformation and fracture of Gilsocarbon graphite subject to service environments

NASA Astrophysics Data System (ADS)

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

2018-02-01

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

Pyrolytic graphite gauge for measuring heat flux

NASA Technical Reports Server (NTRS)

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.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ganesh, Panchapakesan; Kent, Paul R; Mochalin, Vadym N

We simulate the experimentally observed graphitization of nanodiamonds into multi-shell onion-like carbon nanostructures, also called carbon onions, at different temperatures, using reactive force fields. The simulations include long-range Coulomb and van der Waals interactions. Our results suggest that long-range interactions play a crucial role in the phase-stability and the graphitization process. Graphitization is both enthalpically and entropically driven and can hence be controlled with temperature. The outer layers of the nanodiamond have a lower kinetic barrier toward graphitization irrespective of the size of the nanodiamond and graphitize within a few-hundred picoseconds, with a large volume increase. The inner core ofmore » the nanodiamonds displays a large size-dependent kinetic barrier, and graphitizes much more slowly with abrupt jumps in the internal energy. It eventually graphitizes by releasing pressure and expands once the outer shells have graphitized. The degree of transformation at a particular temperature is thereby determined by a delicate balance between the thermal energy, long-range interactions, and the entropic/enthalpic free energy gained by graphitization. Upon full graphitization, a multi-shell carbon nanostructure appears, with a shell-shell spacing of about {approx}3.4 {angstrom} for all sizes. The shells are highly defective with predominantly five- and seven-membered rings to curve space. Larger nanodiamonds with a diameter of 4 nm can graphitize into spiral structures with a large ({approx}29-atom carbon ring) pore opening on the outermost shell. Such a large one-way channel is most attractive for a controlled insertion of molecules/ions such as Li ions, water, or ionic liquids, for increased electrochemical capacitor or battery electrode applications.« less

The graphite deposit at Borrowdale (UK): A catastrophic mineralizing event associated with Ordovician magmatism

NASA Astrophysics Data System (ADS)

Ortega, L.; Millward, D.; Luque, F. J.; Barrenechea, J. F.; Beyssac, O.; Huizenga, J.-M.; Rodas, M.; Clarke, S. M.

2010-04-01

The volcanic-hosted graphite deposit at Borrowdale in Cumbria, UK, was formed through precipitation from C-O-H fluids. The δ 13C data indicate that carbon was incorporated into the mineralizing fluids by assimilation of carbonaceous metapelites of the Skiddaw Group by andesite magmas of the Borrowdale Volcanic Group. The graphite mineralization occurred as the fluids migrated upwards through normal conjugate fractures forming the main subvertical pipe-like bodies. The mineralizing fluids evolved from CO 2-CH 4-H 2O mixtures (XCO 2 = 0.6-0.8) to CH 4-H 2O mixtures. Coevally with graphite deposition, the andesite and dioritic wall rocks adjacent to the veins were intensely hydrothermally altered to a propylitic assemblage. The initial graphite precipitation was probably triggered by the earliest hydration reactions in the volcanic host rocks. During the main mineralization stage, graphite precipitated along the pipe-like bodies due to CO 2 → C + O 2. This agrees with the isotopic data which indicate that the first graphite morphologies crystallizing from the fluid (cryptocrystalline aggregates) are isotopically lighter than those crystallizing later (flakes). Late chlorite-graphite veins were formed from CH 4-enriched fluids following the reaction CH 4 + O 2 → C + 2H 2O, producing the successive precipitation of isotopically lighter graphite morphologies. Thus, as mineralization proceeded, water-generating reactions were involved in graphite precipitation, further favouring the propylitic alteration. The structural features of the pipe-like mineralized bodies as well as the isotopic homogeneity of graphite suggest that the mineralization occurred in a very short period of time.

Solid Fuel Burning in Steady, Strained, Premixed Flow Fields: The Graphite/Air/Methane System

NASA Technical Reports Server (NTRS)

Egolfopoulos, Fokion N.; Wu, Ming-Shin (Technical Monitor)

2000-01-01

A detailed numerical investigation was conducted on the simultaneous burning of laminar premixed CH4/air flames and solid graphite in a stagnation flow configuration. The graphite and methane were chosen for this model, given that they are practical fuels and their chemical kinetics are considered as the most reliable ones among solid and hydrocarbon fuels, respectively. The simulation was performed by solving the quasi-one-dimensional equations of mass, momentum, energy, and species. The GRI 2.1 scheme was used for the gas-phase kinetics, while the heterogeneous kinetics were described by a six-step mechanism including stable and radical species. The effects of the graphite surface temperature, the gas-phase equivalence ratio, and the aerodynamic strain rate on the graphite burning rate and NO, production and destruction mechanisms were assessed. Results indicate that as the graphite temperature increases, its burning rate as well as the NO, concentration increase. Furthermore, it was found that by increasing the strain rate, the graphite burning rate increases as a result of the augmented supply of the gas-phase reactants towards the surface, while the NO, concentration decreases as a result of the reduced residence time. The effect of the equivalence ratio on both the graphite burning rate and NO, concentration was found to be non-monotonic and strongly dependent on the graphite temperature. Comparisons between results obtained for a graphite and a chemically inert surface revealed that the chemical activity of the graphite surface can result to the reduction of NO through reactions of the CH3, CH2, CH, and N radicals with NO.

Reaction rates of graphite with ozone measured by etch decoration

NASA Technical Reports Server (NTRS)

Hennig, G. R.; Montet, G. L.

1968-01-01

Etch-decoration technique of detecting vacancies in graphite has been used to determine the reaction rates of graphite with ozone in the directions parallel and perpendicular to the layer planes. It consists essentially of peeling single atom layers off graphite crystals without affecting the remainder of the crystal.

Graphite-based photovoltaic cells

DOEpatents

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.

78 FR 56864 - Small Diameter Graphite Electrodes From the People's Republic of China: Affirmative Final...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-16

... DEPARTMENT OF COMMERCE International Trade Administration [A-570-929] Small Diameter Graphite... (Department) determines that imports from the People's Republic of China (PRC) of graphite electrodes... Act of 1930, as amended (the Act).\\1\\ \\1\\ See Antidumping Duty Order: Small Diameter Graphite...

New Occurrence of Shocked Graphite Aggregates at Barringer Crater

NASA Astrophysics Data System (ADS)

Miura, Y.; Noma, Y.; Iancu, O. G.

1993-07-01

High-pressure carbon minera]s are considered to be formed by solid-solid transformation under static or impact high-pressure condition, but shocked quartz aggregates of impact craters are considered to be formed by quenched accretion of various aggregates by dynamic impact process [1-3]. The main purpose of this study is to elucidate new findings and occurrences of shocked graphite (SG) aggregates [2,3] at the Barringer meteorite crater. The graphite nodule block of Barringer Crater used in this study is collected near the rim. The sample is compared with standard graphite samples of Korea, Madagascar, and artificial impact graphites. There are four different mineral aggregates of the Barringer graphite nodule sample: (1) shocked graphite-1, (2) shocked graphite-2 and hexagonal diamond in the vein, (3) shocked quartz-1 (with kamacite) in the rim, and (4) calcite in the rim (Table 1). X-ray diffraction peaks of shocked graphite reveal low X-ray intensity, high Bragg-angle shift of X-ray diffraction peak, and multiple splitting of X-ray diffraction peaks. X-ray calculated density (rho) has been determined by X-ray diffractometer by the equation of density deviation Delta rho (%) = 100 x {(rho-rho(sub)0)/rho(sub)0}, where standard density rho(sub)0 is 2.255 g/cm^3 in Korean graphite [2,3]. The high-density value of shocked graphite grain obtained in Barringer is Delta rho = +0.6 +/- 0.1%. Shocked hexagonal diamonds (chaoite) show a high value of Delta rho = +0.6 +/- 0.9%. Analytical electron microscopy data reveal three different aggregates in the graphite nodule samples (Table 1): (1) shocked graphite-1 in the matrix, which contains uniformly Fe and Ca elements formed under gas state; (2) shocked graphite-2 in the vein, where crystallized shocked graphites and hexagonal diamonds are surrounded by kamacite-rich metals formed under gas-melt states of mixed compositions from iron meteorite and target rocks; and (3) shocked quartz-1 and kamacite in the rim, where coexisted elements are supplied from kamacite, sandstone, and limestone. The shocked quartz-1 grains with high density contain Fe and Ca elements that are different from the shocked quartz-2 of pure silica [1] formed at the final stage from the Coconino sandstone. (4) Limestone in the rim is attached from Kaibab limestone. The present shocked graphites with high density are the same as artificial fine-grained shocked graphites (Delta rho = +0.7%). Table 1, which appears here in the hard copy, shows formation stages with two shocked graphites in the Barringer Crater. Formation of shocked aggregates with chemical contamination indicate dynamic accretion processes of quenching and depression at impact. The existence of two shocked graphites indicates the two formation stages of the first gas-state and the second gas-melt states with quenching processes. The origin of carbon in the shocked graphites is considered in this study to be from Kaibab limestone. References: [1] Miura Y. (1991) Shock Waves, 1, 35-41. [2] Miura Y. (1992) Proc. Shock Waves (Japan), 2, 54-57. [3] Miura Y. et al. (1993) Symp. NIPR Antarctic Meteorite (Tokyo), in press. [4] Foote A. E. (1891) Am. J. Sci., 42, 413-417. [5] Hannemann R. E. et al. (1967) Science, 155, 995-997.

Strategic disruption of nuclear pores structure, integrity and barrier for nuclear apoptosis.

PubMed

Shahin, Victor

2017-08-01

Apoptosis is a programmed cell death playing key roles in physiology and pathophysiology of multi cellular organisms. Its nuclear manifestation requires transmission of the death signals across the nuclear pore complexes (NPCs). In strategic sequential steps apoptotic factors disrupt NPCs structure, integrity and barrier ultimately leading to nuclear breakdown. The present review reflects on these steps. Copyright © 2017 Elsevier Ltd. All rights reserved.

Low temperature vapor phase digestion of graphite

DOE Office of Scientific and Technical Information (OSTI.GOV)

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.

AC induction field heating of graphite foam

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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.

PROCESS OF COATING GRAPHITE WITH NIOBIUM-TITANIUM CARBIDE

DOEpatents

Halden, F.A.; Smiley, W.D.; Hruz, F.M.

1961-07-01

A process of coating graphite with niobium - titanium carbide is described. It is found that the addition of more than ten percent by weight of titanium to niobium results in much greater wetting of the graphite by the niobium and a much more adherent coating. The preferred embodiment comprises contacting the graphite with a powdered alloy or mixture, degassing simultaneously the powder and the graphite, and then heating them to a high temperature to cause melting, wetting, spreading, and carburization of the niobium-titanium powder.

A cryogenic thermal source for detector array characterization

NASA Astrophysics Data System (ADS)

Chuss, David T.; Rostem, Karwan; Wollack, Edward J.; Berman, Leah; Colazo, Felipe; DeGeorge, Martin; Helson, Kyle; Sagliocca, Marco

2017-10-01

We describe the design, fabrication, and validation of a cryogenically compatible quasioptical thermal source for characterization of detector arrays. The source is constructed using a graphite-loaded epoxy mixture that is molded into a tiled pyramidal structure. The mold is fabricated using a hardened steel template produced via a wire electron discharge machining process. The absorptive mixture is bonded to a copper backplate enabling thermalization of the entire structure and measurement of the source temperature. Measurements indicate that the reflectance of the source is <0.001 across a spectral band extending from 75 to 330 GHz.

A Cryogenic Thermal Source for Detector Array Characterization

NASA Technical Reports Server (NTRS)

Chuss, David T.; Rostem, Karwan; Wollack, Edward J.; Berman, Leah; Colazo, Felipe; DeGeorge, Martin; Helson, Kyle; Sagliocca, Marco

2017-01-01

We describe the design, fabrication, and validation of a cryogenically compatible quasioptical thermal source for characterization of detector arrays. The source is constructed using a graphite-loaded epoxy mixture that is molded into a tiled pyramidal structure. The mold is fabricated using a hardened steel template produced via a wire electron discharge machining process. The absorptive mixture is bonded to a copper backplate enabling thermalization of the entire structure and measurement of the source temperature. Measurements indicate that the reflectance of the source is less than 0.001 across a spectral band extending from 75 to 330 gigahertz.

Ablation and deceleration of mass-driver launched projectiles for space disposal of nuclear wastes

NASA Astrophysics Data System (ADS)

Park, C.; Bowen, S. W.

1981-01-01

The energy cost of launching a projectile containing nuclear waste is two orders of magnitude lower with a mass driver than with a typical rocket system. A mass driver scheme will be feasible, however, only if ablation and deceleration are within certain tolerable limits. It is shown that if a hemisphere-cylinder-shaped projectile protected thermally with a graphite nose is launched vertically to attain a velocity of 17 km/sec at an altitude of 40 km, the mass loss from ablation during atmospheric flight will be less than 0.1 ton, provided the radius of the projectile is under 20 cm and the projectile's mass is of the order of 1 ton. The velocity loss from drag will vary from 0.4 to 30 km/sec, depending on the mass and radius of the projectile, the smaller velocity loss corresponding to large mass and small radius. Ablation is always within a tolerable range for schemes using a mass driver launcher to dispose of nuclear wastes outside the solar system. Deceleration can also be held in the tolerable range if the mass and diameter of the projectile are properly chosen.

Nuclear thermal propulsion engine system design analysis code development

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.; Ivanenok, Joseph F.

1992-01-01

A Nuclear Thermal Propulsion (NTP) Engine System Design Analyis Code has recently been developed to characterize key NTP engine system design features. Such a versatile, standalone NTP system performance and engine design code is required to support ongoing and future engine system and vehicle design efforts associated with proposed Space Exploration Initiative (SEI) missions of interest. Key areas of interest in the engine system modeling effort were the reactor, shielding, and inclusion of an engine multi-redundant propellant pump feed system design option. A solid-core nuclear thermal reactor and internal shielding code model was developed to estimate the reactor's thermal-hydraulic and physical parameters based on a prescribed thermal output which was integrated into a state-of-the-art engine system design model. The reactor code module has the capability to model graphite, composite, or carbide fuels. Key output from the model consists of reactor parameters such as thermal power, pressure drop, thermal profile, and heat generation in cooled structures (reflector, shield, and core supports), as well as the engine system parameters such as weight, dimensions, pressures, temperatures, mass flows, and performance. The model's overall analysis methodology and its key assumptions and capabilities are summarized in this paper.

An Overview of the Nuclear Electric Xenon Ion System (NEXIS) Activity

NASA Technical Reports Server (NTRS)

Randolph, Thomas M.; Polk, James E., Jr.

2004-01-01

The Nuclear Electric Xenon Ion System (NEXIS) research and development activity within NASA's Project Prometheus, was one of three proposals selected by NASA to develop thruster technologies for long life, high power, high specific impulse nuclear electric propulsion systems that would enable more robust and ambitious science exploration missions to the outer solar system. NEXIS technology represents a dramatic improvement in the state-of-the-art for ion propulsion and is designed to achieve propellant throughput capabilities >= 2000 kg and efficiencies >= 78% while increasing the thruster power to >= 20 kW and specific impulse to >= 6000 s. The NEXIS technology uses erosion resistant carbon-carbon grids, a graphite keeper, a new reservoir hollow cathode, a 65-cm diameter chamber masked to produce a 57-cm diameter ion beam, and a shared neutralizer architecture to achieve these goals. The accomplishments of the NEXIS activity so far include performance testing of a laboratory model thruster, successful completion of a proof of concept reservoir cathode 2000 hour wear test, structural and thermal analysis of a completed development model thruster design, fabrication of most of the development model piece parts, and the nearly complete vacuum facility modifications to allow long duration wear testing of high power ion thrusters.

Late-time particle emission from laser-produced graphite plasma

NASA Astrophysics Data System (ADS)

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

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.

Fire test method for graphite fiber reinforced plastics

NASA Technical Reports Server (NTRS)

Bowles, K. J.

1980-01-01

A potential problem in the use of graphite fiber reinforced resin matrix composites is the dispersal of graphite fibers during accidental fires. Airborne, electrically conductive fibers originating from the burning composites could enter and cause shorting in electrical equipment located in surrounding areas. A test method for assessing the burning characteristics of graphite fiber reinforced composites and the effectiveness of the composites in retaining the graphite fibers has been developed. The method utilizes a modified Ohio State University Rate of Heat Release apparatus. The equipment and the testing procedure are described. The application of the test method to the assessment of composite materials is illustrated for two resin matrix/graphite composite systems.

Square lattice honeycomb tri-carbide fuels for 50 to 250 KN variable thrust NTP design

NASA Astrophysics Data System (ADS)

Anghaie, Samim; Knight, Travis; Gouw, Reza; Furman, Eric

2001-02-01

Ultrahigh temperature solid solution of tri-carbide fuels are used to design an ultracompact nuclear thermal rocket generating 950 seconds of specific impulse with scalable thrust level in range of 50 to 250 kilo Newtons. Solid solutions of tri-carbide nuclear fuels such as uranium-zirconium-niobium carbide. UZrNbC, are processed to contain certain mixing ratio between uranium carbide and two stabilizing carbides. Zirconium or niobium in the tri-carbide could be replaced by tantalum or hafnium to provide higher chemical stability in hot hydrogen environment or to provide different nuclear design characteristics. Recent studies have demonstrated the chemical compatibility of tri-carbide fuels with hydrogen propellant for a few to tens of hours of operation at temperatures ranging from 2800 K to 3300 K, respectively. Fuel elements are fabricated from thin tri-carbide wafers that are grooved and locked into a square-lattice honeycomb (SLHC) shape. The hockey puck shaped SLHC fuel elements are stacked up in a grooved graphite tube to form a SLHC fuel assembly. A total of 18 fuel assemblies are arranged circumferentially to form two concentric rings of fuel assemblies with zirconium hydride filling the space between assemblies. For 50 to 250 kilo Newtons thrust operations, the reactor diameter and length including reflectors are 57 cm and 60 cm, respectively. Results of the nuclear design and thermal fluid analyses of the SLHC nuclear thermal propulsion system are presented. .

Design and development of high efficiency 140W space TWT with graphite collector

NASA Astrophysics Data System (ADS)

Srivastava, V.; Purohit, G.; Sharma, R. K.; Sharma, S. M.; Bera, A.; Bhaskar, P. V.; Singh, R. R.; Prasad, K.; Kiran, V.

2008-05-01

4-stage graphite collector assembly has been designed and developed for a 140W Ku-band space TWT to achieve the collector efficiency more than 80%. The UHV compatible, high density, copper impregnated POCO graphite (DFP-1C) was used to fabricate the four collector electrodes of the 4-stage depressed collector. Copper impregnated graphite material is used for the collector electrodes because of its low secondary electron emission coefficient, high thermal and electrical conductivities, easy machining and brazing, low thermal expansion coefficient and low weight. The graphite material was characterized for the UHV compatibility. The collector electrodes were precisely fabricated by careful machining, and technology was developed for brazing of graphite electrodes with high voltage alumina insulators. Complete TWT with four-stage graphite collector was developed and 140W output power at gain more than 55 dB was achieved. The TWT was pumped from both the gun and the collector ends.

Development of design data for graphite reinforced epoxy and polyimide composites

NASA Technical Reports Server (NTRS)

Scheck, W. G.

1974-01-01

Processing techniques and design data were characterized for a graphite/epoxy composite system that is useful from 75 K to 450 K, and a graphite/polyimide composite system that is useful from 75 K to 589 K. The Monsanto 710 polyimide resin was selected as the resin to be characterized and used with the graphite fiber reinforcement. Material was purchased using the prepreg specification for the design data generation for both the HT-S/710 and HM-S/710 graphite/polyimide composite system. Lamina and laminate properties were determined at 75 K, 297 K, and 589 K. The test results obtained on the skin-stringer components proved that graphite/polyimide composites can be reliably designed and analyzed much like graphite/epoxy composites. The design data generated in the program includes the standard static mechanical properties, biaxial strain data, creep, fatigue, aging, and thick laminate data.

Chemical stabilization of graphite surfaces

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bistrika, Alexander A.; Lerner, Michael M.

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 conditionsmore » 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.« less

Development of CIP/graphite composite additives for electromagnetic wave absorption applications

NASA Astrophysics Data System (ADS)

Woo, Soobin; Yoo, Chan-Sei; Kim, Hwijun; Lee, Mijung; Quevedo-Lopez, Manuel; Choi, Hyunjoo

2017-09-01

In this study, the electromagnetic (EM) wave absorption ability of carbonyl iron powder (CIP)/graphite composites produced by ball milling were studied in a range of 28.5 GHz to examine the effects of the morphology and volume fraction of graphite on EM wave absorption ability. The results indicated that a ball milling technique was effective in exfoliating the graphite and covering it with CIP, thereby markedly increasing the specific surface area of the hybrid powder. The increase in the surface area and hybridization with dielectric loss materials (i.e., graphite) improved EM absorbing properties of CIP in the range of S and X bands. Specifically, the CIP/graphite composite containing 3 wt% graphite exhibited electromagnetic wave absorption of -13 dB at 7 GHz, -21 dB at 5.8 GHz, and -29 dB at 4.3 GHz after 1 h, 8 h, and 16 h of milling, respectively. [Figure not available: see fulltext.

Friction and wear of metals in contact with pyrolytic graphite

NASA Technical Reports Server (NTRS)

Buckley, D. H.; Brainard, W. A.

1975-01-01

Sliding friction experiments were conducted with gold, iron, and tantalum single crystals sliding on prismatic and basal orientations of pyrolytic graphite in various environments, including vacuum, oxygen, water vapor, nitrogen, and hydrogen bromide. Surfaces were examined in the clean state and with various adsorbates present on the graphite surfaces. Auger and LEED spectroscopy, SEM, and EDXA were used to characterize the graphite surfaces. Results indicate that the prismatic and basal orientations do not contain nor do they chemisorb oxygen, water vapor, acetylene, or hydrogen bromide. All three metals exhibited higher friction on the prismatic than on the basal orientation and these metals transferred to the atomically clean prismatic orientation of pyrolytic graphite. No metal transfer to the graphite was observed in the presence of adsorbates at 760 torr. Ion bombardment of the graphite surface with nitrogen ions resulted in the adherence of nitrogen to the surface.

Formation mechanism of graphite hexagonal pyramids by argon plasma etching of graphite substrates

NASA Astrophysics Data System (ADS)

Glad, X.; de Poucques, L.; Bougdira, J.

2015-12-01

A new graphite crystal morphology has been recently reported, namely the graphite hexagonal pyramids (GHPs). They are hexagonally-shaped crystals with diameters ranging from 50 to 800 nm and a constant apex angle of 40°. These nanostructures are formed from graphite substrates (flexible graphite and highly ordered pyrolytic graphite) in low pressure helicon coupling radiofrequency argon plasma at 25 eV ion energy and, purportedly, due to a physical etching process. In this paper, the occurrence of peculiar crystals is shown, presenting two hexagonal orientations obtained on both types of samples, which confirms such a formation mechanism. Moreover, by applying a pretreatment step with different time durations of inductive coupling radiofrequency argon plasma, for which the incident ion energy decreases at 12 eV, uniform coverage of the surface can be achieved with an influence on the density and size of the GHPs.

Tuning graphitic oxide for initiator- and metal-free aerobic epoxidation of linear alkenes

NASA Astrophysics Data System (ADS)

Pattisson, Samuel; Nowicka, Ewa; Gupta, Upendra N.; Shaw, Greg; Jenkins, Robert L.; Morgan, David J.; Knight, David W.; Hutchings, Graham J.

2016-09-01

Graphitic oxide has potential as a carbocatalyst for a wide range of reactions. Interest in this material has risen enormously due to it being a precursor to graphene via the chemical oxidation of graphite. Despite some studies suggesting that the chosen method of graphite oxidation can influence the physical properties of the graphitic oxide, the preparation method and extent of oxidation remain unresolved for catalytic applications. Here we show that tuning the graphitic oxide surface can be achieved by varying the amount and type of oxidant. The resulting materials differ in level of oxidation, surface oxygen content and functionality. Most importantly, we show that these graphitic oxide materials are active as unique carbocatalysts for low-temperature aerobic epoxidation of linear alkenes in the absence of initiator or metal. An optimum level of oxidation is necessary and materials produced via conventional permanganate-based methods are far from optimal.

Construction and performance evaluation of mediator-less microbial fuel cell using carbon nanotubes as an anode material.

PubMed

Roh, Sung-Hee; Kim, Sun-Il

2012-05-01

A microbial fuel cell (MFC) is a device that converts chemical energy to electrical energy using the catalytic reaction of microorganisms. We investigated the performance of mediator-less MFC with carbon nanotubes (CNTs)/graphite felt composite electrodes. The addition of CNTs to a graphite felt electrode increases the specific surface area of the electrode and enhances the charge transfer capability so as to cause considerable improvement of the electrochemical activity for the anode reaction in a MFC. The performance of the MFC using CNTs/graphite felt electrode has been compared against a plain graphite felt electrode based MFC. A CNTs/graphite felt electrode showed as high as 15% increase in the power density (252 mW/m2) compared to graphite felt electrode (214 mW/m2). The CNTs/graphite felt anode therefore offers good prospects for application in MFCs.

Method of making segmented pyrolytic graphite sputtering targets

DOEpatents

McKernan, Mark A.; Alford, Craig S.; Makowiecki, Daniel M.; Chen, Chih-Wen

1994-01-01

Anisotropic pyrolytic graphite wafers are oriented and bonded together such that the graphite's high thermal conductivity planes are maximized along the back surface of the segmented pyrolytic graphite target to allow for optimum heat conduction away from the sputter target's sputtering surface and to allow for maximum energy transmission from the target's sputtering surface.

76 FR 36092 - Small Diameter Graphite Electrodes From the People's Republic of China: Extension of Time Limit...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-21

... DEPARTMENT OF COMMERCE International Trade Administration [A-570-929] Small Diameter Graphite... antidumping duty order on small diameter graphite electrodes from the People's Republic of China (``PRC'') for... preliminary results of this review were published on March 7, 2011. See Small Diameter Graphite Electrodes...

77 FR 6060 - Small Diameter Graphite Electrodes from the People's Republic of China: Extension of Time Limit...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-07

... DEPARTMENT OF COMMERCE International Trade Administration [A-570-929] Small Diameter Graphite... Department) initiated an administrative review of the antidumping duty order on small diameter graphite... preliminary results of this review by 95 days until February 3, 2012. See Small Diameter Graphite Electrodes...

76 FR 67411 - Small Diameter Graphite Electrodes From the People's Republic of China: Extension of Time Limit...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-01

... DEPARTMENT OF COMMERCE International Trade Administration [A-570-929] Small Diameter Graphite... diameter graphite electrodes from the People's Republic of China (PRC) for the period February 1, 2010... Graphite, Co. The preliminary results of the review are currently due no later than October 31, 2011...

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

NASA Astrophysics Data System (ADS)

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

2018-04-01

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