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.

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.

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.

Sealing nuclear graphite with pyrolytic carbon

NASA Astrophysics Data System (ADS)

Feng, Shanglei; Xu, Li; Li, Li; Bai, Shuo; Yang, Xinmei; Zhou, Xingtai

2013-10-01

Pyrolytic carbon (PyC) coatings were deposited on IG-110 nuclear graphite by thermal decomposition of methane at ∼1830 °C. The PyC coatings are anisotropic and airtight enough to protect IG-110 nuclear graphite against the permeation of molten fluoride salts and the diffusion of gases. The investigations indicate that the sealing nuclear graphite with PyC coating is a promising method for its application in Molten Salt Reactor (MSR).

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

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.

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

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.

Comparison between the Strength Levels of Baseline Nuclear-Grade Graphite and Graphite Irradiated in AGC-2

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

Carroll, Mark Christopher

2015-07-01

This report details the initial comparison of mechanical strength properties between the cylindrical nuclear-grade graphite specimens irradiated in the second Advanced Graphite Creep (AGC-2) experiment with the established baseline, or unirradiated, mechanical properties compiled in the Baseline Graphite Characterization program. The overall comparative analysis will describe the development of an appropriate test protocol for irradiated specimens, the execution of the mechanical tests on the AGC-2 sample population, and will further discuss the data in terms of developing an accurate irradiated property distribution in the limited amount of irradiated data by leveraging the considerably larger property datasets being captured in themore » Baseline Graphite Characterization program. Integrating information on the inherent variability in nuclear-grade graphite with more complete datasets is one of the goals of the VHTR Graphite Materials program. Between “sister” specimens, or specimens with the same geometry machined from the same sub-block of graphite from which the irradiated AGC specimens were extracted, and the Baseline datasets, a comprehensive body of data will exist that can provide both a direct and indirect indication of the full irradiated property distributions that can be expected of irradiated nuclear-grade graphite while in service in a VHTR system. While the most critical data will remain the actual irradiated property measurements, expansion of this data into accurate distributions based on the inherent variability in graphite properties will be a crucial step in qualifying graphite for nuclear use as a structural material in a VHTR environment.« less

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.

Thermodynamic Simulation of Equilibrium Composition of Reaction Products at Dehydration of a Technological Channel in a Uranium-Graphite Reactor

NASA Astrophysics Data System (ADS)

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

2018-01-01

The problems of accumulation of nuclear fuel spills in the graphite stack in the course of operation of uranium-graphite nuclear reactors are considered. The results of thermodynamic analysis of the processes in the graphite stack at dehydration of a technological channel, fuel element shell unsealing and migration of fission products, and activation of stable nuclides in structural elements of the reactor and actinides inside the graphite moderator are given. The main chemical reactions and compounds that are produced in these modes in the reactor channel during its operation and that may be hazardous after its shutdown and decommissioning are presented. Thermodynamic simulation of the equilibrium composition is performed using the specialized code TERRA. The results of thermodynamic simulation of the equilibrium composition in different cases of technological channel dehydration in the course of the reactor operation show that, if the temperature inside the active core of the nuclear reactor increases to the melting temperature of the fuel element, oxides and carbides of nuclear fuel are produced. The mathematical model of the nonstationary heat transfer in a graphite stack of a uranium-graphite reactor in the case of the technological channel dehydration is presented. The results of calculated temperature evolution at the center of the fuel element, the replaceable graphite element, the air gap, and in the surface layer of the block graphite are given. The numerical results show that, in the case of dehydration of the technological channel in the uranium-graphite reactor with metallic uranium, the main reaction product is uranium dioxide UO2 in the condensed phase. Low probability of production of pyrophoric uranium compounds (UH3) in the graphite stack is proven, which allows one to disassemble the graphite stack without the risk of spontaneous graphite ignition in the course of decommissioning of the uranium-graphite nuclear reactor.

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

The Fracture Toughness of Nuclear Graphites Grades

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

Burchell, Timothy D.; Erdman, III, Donald L.; Lowden, Rick R.

2017-04-01

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-24

... NUCLEAR REGULATORY COMMISSION Request for a License To Export Nuclear Grade Graphite Pursuant to 10 CFR 110.70 (b) ``Public Notice of Receipt of an Application,'' please take notice that the Nuclear... requestor or petitioner upon the applicant, the office of the General Counsel, U.S. Nuclear Regulatory...

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

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

Spicer, James

Among techniques that have been used to determine elastic modulus in nuclear graphites, ultrasonic methods have enjoyed wide use and standards using contacting piezoelectric tranducers have been developed to ensure repeatability of these types of measurements. However, the use of couplants and the pressures used to effectively couple transducers to samples can bias measurements and produce results that are not wholly related to the properties of the graphite itself. In this work, we have investigated the use of laser ultrasonic methods for making elastic modulus measurements in nuclear graphites. These methods use laser-based transmitters and receivers to gather data andmore » do not require use of ultrasonic couplants or mechanical contact with the sample. As a result, information directly related to the elastic responses of graphite can be gathered even if the graphite is porous, brittle and compliant. In particular, we have demonstrated the use of laser ultrasonics for the determination of both Young’s modulus and shear modulus in a range of nuclear graphites including those that are being considered for use in future nuclear reactors. These results have been analyzed to assess the contributions of porosity and microcracking to the elastic responses of these graphites. Laser-based methods have also been used to assess the moduli of NBG-18 and IG-110 where samples of each grade were oxidized to produce specific changes in porosity. These data were used to develop new models for the elastic responses of nuclear graphites and these models have been used to infer specific changes in graphite microstructure that occur during oxidation that affect elastic modulus. Specifically, we show how ultrasonic measurements in oxidized graphites are consistent with nano/microscale oxidation processes where basal plane edges react more readily than basal plane surfaces. We have also shown the use of laser-based methods to perform shear-wave birefringence measurements and have shown how these measurements can be used to assess elastic anisotropy in nuclear graphites. Using models developed in this program, ultrasonic data were interpreted to extract orientation distribution coefficients that could be used to represent anisotropy in these materials. This demonstration showed the use of ultrasonic methods to quantify anisotropy and how these methods provide more detailed information than do measurements of thermal expansion – a technique commonly used for assessing anisotropy in nuclear graphites. Finally, we have employed laser-based, ultrasonic-correlation techniques in attempts to quantify aspects of graphite microstructure such as pore size and distribution. Results of these measurements indicate that additional work must be performed to make this ultrasonic approach viable for quantitative microstructural characterization.« 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

METHOD OF FABRICATING A GRAPHITE MODERATED REACTOR

DOEpatents

Kratz, H.R.

1963-05-01

S>A nuclear reactor formed of spaced bodies of uranium and graphite blocks is improved by diffusing helium through the graphite blocks in order to replace the air in the pores of the graphite with helium. The helium-impregnated graphite conducts heat better, and absorbs neutrons less, than the original air- impregnated graphite. (AEC)

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

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 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

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.

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.

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.

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

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.

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.

Method of producing exfoliated graphite composite compositions for fuel cell flow field plates

DOEpatents

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

2014-04-08

A method of producing an electrically conductive composite composition, which is particularly useful for fuel cell bipolar plate applications. The method comprises: (a) providing a supply of expandable graphite powder; (b) providing a supply of a non-expandable powder component comprising a binder or matrix material; (c) blending the expandable graphite with the non-expandable powder component to form a powder mixture wherein the non-expandable powder component is in the amount of between 3% and 60% by weight based on the total weight of the powder mixture; (d) exposing the powder mixture to a temperature sufficient for exfoliating the expandable graphite to obtain a compressible mixture comprising expanded graphite worms and the non-expandable component; (e) compressing the compressible mixture at a pressure within the range of from about 5 psi to about 50,000 psi in predetermined directions into predetermined forms of cohered graphite composite compact; and (f) treating the so-formed cohered graphite composite to activate the binder or matrix material thereby promoting adhesion within the compact to produce the desired composite composition. Preferably, the non-expandable powder component further comprises an isotropy-promoting agent such as non-expandable graphite particles. Further preferably, step (e) comprises compressing the mixture in at least two directions. The method leads to composite plates with exceptionally high thickness-direction electrical conductivity.

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.

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.

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 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.

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.

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

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

PubMed

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

2015-01-01

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

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

PubMed Central

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

2015-01-01

Pile grade A (PGA) graphite was used as a material for moderating and reflecting neutrons in the UK’s first generation Magnox nuclear power reactors. As all but one of these reactors are now shut down there is a need to understand the residual state of the material prior to decommissioning of the cores, in particular the location and concentration of key radio-contaminants such as 14C. The oxidation behaviour of unirradiated PGA graphite was studied, in the temperature range 600–1050°C, in air and nitrogen using thermogravimetric analysis, scanning electron microscopy and X-ray tomography to investigate the possibility of using thermal degradation techniques to examine 14C distribution within irradiated material. The thermal decomposition of PGA graphite was observed to follow the three oxidation regimes historically identified by previous workers with limited, uniform oxidation at temperatures below 600°C and substantial, external oxidation at higher temperatures. This work demonstrates that the different oxidation regimes of PGA graphite could be developed into a methodology to characterise the distribution and concentration of 14C in irradiated graphite by thermal treatment. PMID:26575374

Characterization of PMR-15 polyimide composition in thermo-oxidatively exposed graphite fiber composites

NASA Technical Reports Server (NTRS)

Alston, W. B.

1980-01-01

The contributions of individual resin components to total resin weight loss in 600 F air aged Celion 6000/PMR-15 polyimide composites were determined from the overall resin weight loss in the composite by chemically separating the PMR-15 matrix resin into its monomeric components. The individual resin components were also analyzed by spectroscopic techniques in order to elucidate curing and degradation mechanisms of the PMR-15 matrix resin. The isothermal weight loss of the individual resin components during prolonged 600 F thermo-oxidative aging of the composite was correlated to the changes observed in the Fourier Transform infrared spectra and Fourier Transform nuclear magnetic resonance spectra of the individual resin components. The correlation was used to identify the molecular site of the thermo-oxidative changes in PMR-15 polyimide matrix resin during 600 F curing the prolonged 600 F thermo-oxidative aging.

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

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.

Facile Synthesis of Pre-Doping Lithium-Ion Into Nitrogen-Doped Graphite Negative Electrode for Lithium-Ion Capacitor.

PubMed

Lee, Seul-Yi; Kim, Ji-Il; Rhee, Kyong Yop; Park, Soo-Jin

2015-09-01

Nitrogen-doped graphite, prepared via the thermal decomposition of melamine into a carbon matrix for use as the negative electrode in lithium-ion capacitors (LICs), was evaluated by electrochemical measurements. Furthermore, in order to study the performance of pre-doped lithium components as a function of nitrogen-doped material, the pre-doped lithium graphite was allowed to react with a lithium salt solution. The results showed that the nitrogen functional groups in the graphite largely influenced the pre-doped lithium components, thereby contributing to the discharge capacity and cycling performance. We confirmed that the large initial irreversible capacity could be significantly decreased by using pre-doped lithium components obtained through the nitrogen-doping method.

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.

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.

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

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.

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

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.

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.

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

Casting of weldable graphite/magnesium metal matrix composites with built-in metallic inserts

NASA Technical Reports Server (NTRS)

Lee, Jonathan A.; Kashalikar, Uday; Majkowski, Patricia

1994-01-01

Technology innovations directed at the advanced development of a potentially low cost and weldable graphite/magnesium metal matrix composites (MMC) through near net shape pressure casting are described. These MMC components uniquely have built-in metallic inserts to provide an innovative approach for joining or connecting other MMC components through conventional joining techniques such as welding, brazing, mechanical fasteners, etc. Moreover, the metallic inserts trapped within the MMC components can be made to transfer the imposed load efficiently to the continuous graphite fiber reinforcement thus producing stronger, stiffer, and more reliable MMC components. The use of low pressure near net shape casting is economical compared to other MMC fabrication processes. These castable and potentially weldable MMC components can provide great payoffs in terms of high strength, high stiffness, low thermal expansion, lightweight, and easily joinable MMC components for several future NASA space structural, industrial, and commercial applications.

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

NASA Astrophysics Data System (ADS)

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

2009-04-01

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

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.

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

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

Flight service evaluation of an advanced composite empennage component on commercial transport aircraft

NASA Technical Reports Server (NTRS)

1976-01-01

The development and flight evaluation of an advanced composite empennage component is presented. The recommended concept for the covers is graphite-epoxy hats bonded to a graphite-epoxy skin. The hat flare-out has been eliminated, instead the hat is continuous into the joint. The recommended concept for the spars is graphite-epoxy caps and a hybrid of Kevlar-49 and graphite-epoxy in the spar web. The spar cap, spar web stiffeners for attaching the ribs, and intermediate stiffeners are planned to be fabricated as a unit. Access hole in the web will be reinforced with a donut type, zero degree graphite-epoxy wound reinforcement. The miniwich design concept in the upper three ribs originally proposed is changed to a graphite-epoxy stiffened solid laminate design concept. The recommended configuration for the lower seven ribs remains as graphite-epoxy caps with aluminum cruciform diagonals. The indicated weight saving for the current advanced composite vertical fin configuration is 20.2% including a 24 lb growth allowance. The project production cost saving is approximately 1% based on a cumulative average of 250 aircraft and including only material, production labor, and quality assurance costs.

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

Flight service evaluation of composite helicopter components

NASA Technical Reports Server (NTRS)

Rich, M. J.; Lowry, D. W.

1983-01-01

This first interim report presents the technical background for including environmental effects in the design of helicopter composite structures, and test results after approximately two year field exposure of components and panels. Composite structural components were removed from Sikorsky S-76 helicopters commercially operated in the Gulf Coast region of Louisiana. Fatigue tests were conducted for a graphite/epoxy tail rotor spar and static test for a graphite/epoxy and Kevlar/epoxy stabilizer. Graphite/epoxy and Kevlar/epoxy panels are being exposed to the outdoor environment in Stratford, Connecticut and West Palm Beach, Florida. For this reporting period the two year panels were returned, moisture measurements taken, and strength tests conducted. Results are compared with initial type certificate strengths for components and with initial laboratory coupon tests for the exposed panels. Comparisons are also presented with predicted and measured moisture contents.

Flight service evaluation of composite helicopter components

NASA Technical Reports Server (NTRS)

Rich, M. J.; Lowry, D. W.

1982-01-01

This first interim report presents the technical background for including environmental effects in the design of helicopter composite structures, and test results after approximately two year field exposure of components and panels. Composite structural components were removed from Sikorsky S-76 helicopters commercially operated in the Gulf Coast region of Louisiana. Fatigue tests were conducted for a graphite/epoxy tail rotor spar and static test for a graphite/epoxy and Kevlar/epoxy stabilizer. Graphite/epoxy and Kevlar/epoxy panels are being exposed to the outdoor environment in Stratford, Connecticut and West Palm Beach, Florida. For this reporting period the two year panels were returned, moisture measurements taken, and strength tests conducted. Results are compared with initial type certificate strengths for components and with initial laboratory coupon tests for the exposed panels. Comparisons are also presented with predicted and measured moisture contents.

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.

Developments in Hollow Graphite Fiber Technology

NASA Technical Reports Server (NTRS)

Stallcup, Michael; Brantley, Lott W., Jr. (Technical Monitor)

2002-01-01

Hollow graphite fibers will be lighter than standard solid graphite fibers and, thus, will save weight in optical components. This program will optimize the processing and properties of hollow carbon fibers developed by MER and to scale-up the processing to produce sufficient fiber for fabricating a large ultra-lightweight mirror for delivery to NASA.

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

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.

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

Analysis of Fission Products on the AGR-1 Capsule Components

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

Paul A. Demkowicz; Jason M. Harp; Philip L. Winston

2013-03-01

The components of the AGR-1 irradiation capsules were analyzed to determine the retained inventory of fission products in order to determine the extent of in-pile fission product release from the fuel compacts. This includes analysis of (i) the metal capsule components, (ii) the graphite fuel holders, (iii) the graphite spacers, and (iv) the gas exit lines. The fission products most prevalent in the components were Ag-110m, Cs 134, Cs 137, Eu-154, and Sr 90, and the most common location was the metal capsule components and the graphite fuel holders. Gamma scanning of the graphite fuel holders was also performed tomore » determine spatial distribution of Ag-110m and radiocesium. Silver was released from the fuel components in significant fractions. The total Ag-110m inventory found in the capsules ranged from 1.2×10 2 (Capsule 3) to 3.8×10 1 (Capsule 6). Ag-110m was not distributed evenly in the graphite fuel holders, but tended to concentrate at the axial ends of the graphite holders in Capsules 1 and 6 (located at the top and bottom of the test train) and near the axial center in Capsules 2, 3, and 5 (in the center of the test train). The Ag-110m further tended to be concentrated around fuel stacks 1 and 3, the two stacks facing the ATR reactor core and location of higher burnup, neutron fluence, and temperatures compared with Stack 2. Detailed correlation of silver release with fuel type and irradiation temperatures is problematic at the capsule level due to the large range of temperatures experienced by individual fuel compacts in each capsule. A comprehensive Ag 110m mass balance for the capsules was performed using measured inventories of individual compacts and the inventory on the capsule components. For most capsules, the mass balance was within 11% of the predicted inventory. The Ag-110m release from individual compacts often exhibited a very large range within a particular capsule.« less

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 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.

Lightweight orthotic appliances

NASA Technical Reports Server (NTRS)

Baucom, R. M.; St. Clair, T. L.

1976-01-01

Graphite-filament reinforced polymer materials are used in applications requiring high tensile strength and modulus. Superior properties of graphite composite materials permit fabrication of supports that are considerably lighter, thinner, and stiffer than conventional components.

A high-performance ternary Si composite anode material with crystal graphite core and amorphous carbon shell

NASA Astrophysics Data System (ADS)

Sui, Dong; Xie, Yuqing; Zhao, Weimin; Zhang, Hongtao; Zhou, Ying; Qin, Xiting; Ma, Yanfeng; Yang, Yong; Chen, Yongsheng

2018-04-01

Si is a promising anode material for lithium-ion batteries, but suffers from sophisticated engineering structures and complex fabrication processes that pose challenges for commercial application. Herein, a ternary Si/graphite/pyrolytic carbon (SiGC) anode material with a structure of crystal core and amorphous shell using low-cost raw materials is developed. In this ternary SiGC composite, Si component exists as nanoparticles and is spread on the surface of the core graphite flakes while the sucrose-derived pyrolytic carbon further covers the graphite/Si components as the amorphous shell. With this structure, Si together with the graphite contributes to the high specific capacity of this Si ternary material. Also the graphite serves as the supporting and conducting matrix and the amorphous shell carbon could accommodate the volume change effect of Si, reinforces the integrity of the composite architecture, and prevents the graphite and Si from direct exposing to the electrolyte. The optimized ternary SiGC composite displays high reversible specific capacity of 818 mAh g-1 at 0.1 A g-1, initial Coulombic efficiency (CE) over 80%, and excellent cycling stability at 0.5 A g-1 with 83.6% capacity retention (∼610 mAh g-1) after 300 cycles.

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.

Status of Chronic Oxidation Studies of Graphite

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

Contescu, Cristian I.; Mee, Robert W.

Graphite will undergo extremely slow, but continuous oxidation by traces of moisture that will be present, albeit at very low levels, in the helium coolant of HTGR. This chronic oxidation may cause degradation of mechanical strength and thermal properties of graphite components if a porous oxidation layer penetrates deep enough in the bulk of graphite components during the lifetime of the reactor. The current research on graphite chronic oxidation is motivated by the acute need to understand the behavior of each graphite grade during prolonged exposure to high temperature chemical attack by moisture. The goal is to provide the elementsmore » needed to develop predictive models for long-time oxidation behavior of graphite components in the cooling helium of HTGR. The tasks derived from this goal are: (1) Oxidation rate measurements in order to determine and validate a comprehensive kinetic model suitable for prediction of intrinsic oxidation rates as a function of temperature and oxidant gas composition; (2) Characterization of effective diffusivity of water vapor in the graphite pore system in order to account for the in-pore transport of moisture; and (3) Development and validation of a predictive model for the penetration depth of the oxidized layer, in order to assess the risk of oxidation caused damage of particular graphite grades after prolonged exposure to the environment of helium coolant in HTGR. The most important and most time consuming of these tasks is the measurement of oxidation rates in accelerated oxidation tests (but still under kinetic control) and the development of a reliable kinetic model. This report summarizes the status of chronic oxidation studies on graphite, and then focuses on model development activities, progress of kinetic measurements, validation of results, and improvement of the kinetic models. Analysis of current and past results obtained with three grades of showed that the classical Langmuir-Hinshelwood model cannot reproduce all data collected so far. Starting from here we propose a modification of the LH model to include temperature activation of graphite surface as a Boltzmann activation function. The enhanced Boltzmann-Langmuir-Hinshelwood model (BLH) was tested successfully on three grades of graphite. The model is a robust, comprehensive mathematical function that allows better fitting of experimental results spanning a wide range of temperature and partial pressures of water vapor and hydrogen. However, the model did not fit satisfactorily the data extracted from the old report on graphite H-451 oxidation by water.« less

A First Look at Graphite Grains from Orgueil: Morphology, Carbon, Nitrogen and Neon Isotopic Compositions of Individual, Chemically Separated Grains

NASA Technical Reports Server (NTRS)

Pravdivtseva, O.; Zinner, E.; Meshik, A. P.; Hohenberg, C. M.; Walker, R. W.

2004-01-01

Presolar graphite in Murchison has been extensively studied. It is characterized by a unique Ne isotopic composition, known as the Ne-E(L) component. According to studies by Huss and Lewis, the concentration of Ne-E(L) in Orgueil is about one order of magnitude higher than in Murchison, when normalized to the matrix. This could be due to a higher presolar graphite abundance in Orgueil, or due to a higher Ne-E concentrations per grain. The Ne isotopic compositions in individual presolar graphite grains from Murchison have been measured before. It was shown, that a third of the grains have detectable excesses in 22Ne, characteristic of the Ne-E(L) component. One grain in a hundred had a Ne-22 concentration two orders of magnitude higher than blank.

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.

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.

Overview of SBIR Phase II Work on Hollow Graphite Fibers

NASA Technical Reports Server (NTRS)

Stallcup, Michael; Brantley, Lott W. (Technical Monitor)

2001-01-01

Ultra-Lightweight materials are enabling for producing space based optical components and support structures. Heretofore, innovative designs using existing materials has been the approach to produce lighter-weight optical systems. Graphite fiber reinforced composites, because of their light weight, have been a material of frequent choice to produce space based optical components. Hollow graphite fibers would be lighter than standard solid graphite fibers and, thus, would save weight in optical components. The Phase I SBIR program demonstrated it is possible to produce hollow carbon fibers that have strengths up to 4.2 GPa which are equivalent to commercial fibers, and composites made from the hollow fibers had substantially equivalent composite strengths as commercial fiber composites at a 46% weight savings. The Phase II SBIR program will optimize processing and properties of the hollow carbon fiber and scale-up processing to produce sufficient fiber for fabricating a large ultra-lightweight mirror for delivery to NASA. Information presented here includes an overview of the strength of some preliminary hollow fibers, photographs of those fibers, and a short discussion of future plans.

Postbuckling behavior of graphite-epoxy panels

NASA Technical Reports Server (NTRS)

Starnes, J. H., Jr.; Dickson, J. N.; Rouse, M.

1984-01-01

Structurally efficient fuselage panels are often designed to allow buckling to occur at applied loads below ultimate. Interest in applying graphite-epoxy materials to fuselage primary structure led to several studies of the post-buckling behavior of graphite-epoxy structural components. Studies of the postbuckling behavior of flat and curved, unstiffened and stiffened graphite-epoxy panels loaded in compression and shear were summarized. The response and failure characteristics of specimens studied experimentally were described, and analytical and experimental results were compared. The specimens tested in the studies described were fabricated from commercially available 0.005-inch-thick unidirectional graphite-fiber tapes preimpregnated with 350 F cure thermosetting epoxy resins.

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

Development and demonstration of manufacturing processes for fabricating graphite/LARC 160 polyimide structural elements

NASA Technical Reports Server (NTRS)

Frost, R. K.; Jones, J. S.; Dynes, P. J.; Wykes, D. H.

1981-01-01

The development and demonstration of manufacturing technologies for the structural application of Celion graphite/LARC-160 polyimide composite material is discussed. Process development and fabrication of demonstration components are discussed. Process development included establishing quality assurance of the basic composite material and processing, nondestructive inspection of fabricated components, developing processes for specific structural forms, and qualification of processes through mechanical testing. Demonstration components were fabricated. The demonstration components consisted of flat laminates, skin/stringer panels, honeycomb panels, chopped fiber compression moldings, and a technology demonstrator segment (TDS) representative of the space shuttle aft body flap.

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.

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.

Design, fabrication and test of graphite/epoxy metering truss structure components, phase 3

NASA Technical Reports Server (NTRS)

1974-01-01

The design, materials, tooling, manufacturing processes, quality control, test procedures, and results associated with the fabrication and test of graphite/epoxy metering truss structure components exhibiting a near zero coefficient of thermal expansion are described. Analytical methods were utilized, with the aid of a computer program, to define the most efficient laminate configurations in terms of thermal behavior and structural requirements. This was followed by an extensive material characterization and selection program, conducted for several graphite/graphite/hybrid laminate systems to obtain experimental data in support of the analytical predictions. Mechanical property tests as well as the coefficient of thermal expansion tests were run on each laminate under study, the results of which were used as the selection criteria for the single most promising laminate. Further coefficient of thermal expansion measurement was successfully performed on three subcomponent tubes utilizing the selected laminate.

Fabrication of graphite/polyimide composite structures.

NASA Technical Reports Server (NTRS)

Varlas, M.

1972-01-01

Selection of graphite/polyimide composite as a prime candidate for high-temperature structural applications involving long-duration temperature environments of 400 to 600 F. A variety of complex graphite/polyimide components has been fabricated, using a match-metal die approach developed for making fiber-reinforced resin composites. Parts produced include sections of a missile adapter skin flange, skin frame section, and I-beam and hat-section stringers, as well as unidirectional (0 deg) and plus or minus 45 deg oriented graphite/polyimide tubes in one-, two-, and six-inch diameters.

Structural changes in a commercial lithium-ion battery during electrochemical cycling: An in situ neutron diffraction study

NASA Astrophysics Data System (ADS)

Sharma, Neeraj; Peterson, Vanessa K.; Elcombe, Margaret M.; Avdeev, Maxim; Studer, Andrew J.; Blagojevic, Ned; Yusoff, Rozila; Kamarulzaman, Norlida

The structural response to electrochemical cycling of the components within a commercial Li-ion battery (LiCoO 2 cathode, graphite anode) is shown through in situ neutron diffraction. Lithuim insertion and extraction is observed in both the cathode and anode. In particular, reversible Li incorporation into both layered and spinel-type LiCoO 2 phases that comprise the cathode is shown and each of these components features several phase transitions attributed to Li content and correlated with the state-of-charge of the battery. At the anode, a constant cell voltage correlates with a stable lithiated graphite phase. Transformation to de-lithiated graphite at the discharged state is characterised by a sharp decrease in both structural cell parameters and cell voltage. In the charged state, a two-phase region exists and is composed of the lithiated graphite phase and about 64% LiC 6. It is postulated that trapping Li in the solid|electrolyte interface layer results in minimal structural changes to the lithiated graphite anode across the constant cell voltage regions of the electrochemical cycle.

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

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.

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.

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.

Technology of civil usage of composites. [in commercial aircraft structures

NASA Technical Reports Server (NTRS)

Kemp, D. E.

1977-01-01

The paper deals with the use of advanced composites in structural components of commercial aircraft. The need for testing the response of a material system to service environment is discussed along with methods for evaluating design and manufacturing aspects of a built-up structure under environmental conditions and fail-safe (damage-tolerance) evaluation of structures. Crashworthiness aspects, the fire-hazard potential, and electrical damage of composite structures are considered. Practical operational experience with commercial aircraft is reviewed for boron/epoxy foreflaps, Kevlar/epoxy fillets and fairings, graphite/epoxy spoilers, graphite/polysulfone spoilers, graphite/epoxy floor posts, boron/aluminum aft pylon skin panels, graphite/epoxy engine nose cowl outer barrels, and graphite/epoxy upper aft rudder segments.

Development of seal ring carbon-graphite materials (tasks 5, 6, and 7)

NASA Technical Reports Server (NTRS)

Fechter, N. J.; Petrunich, P. S.

1972-01-01

Carbon-graphite seal ring bodies for operation at air temperatures to 1300 F(704 C) were manufactured from three select formulations. Mechanical and thermal properties, porosities, and oxidation rates were measured. The results have shown that: (1) Major property improvements anticipated from the screening studies were not realized because of processing problems associated with the scale-up in material size and probable deterioration of a phenolic resin binder; (2) the mechanical properties of a phenolic resin-bonded, carbon-graphite material can be improved by applying high pressure during carbonization; and (3) the textile form of graphite fiber used as the minor filler component in a carbon-graphite material can beneficially affect mechanical properties.

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.

A "Sticky" Mucin-Inspired DNA-Polysaccharide Binder for Silicon and Silicon-Graphite Blended Anodes in Lithium-Ion Batteries.

PubMed

Kim, Sunjin; Jeong, You Kyeong; Wang, Younseon; Lee, Haeshin; Choi, Jang Wook

2018-05-14

New binder concepts have lately demonstrated improvements in the cycle life of high-capacity silicon anodes. Those binder designs adopt adhesive functional groups to enhance affinity with silicon particles and 3D network conformation to secure electrode integrity. However, homogeneous distribution of silicon particles in the presence of a substantial volumetric content of carbonaceous components (i.e., conductive agent, graphite, etc.) is still difficult to achieve while the binder maintains its desired 3D network. Inspired by mucin, the amphiphilic macromolecular lubricant, secreted on the hydrophobic surface of gastrointestine to interface aqueous serous fluid, here, a renatured DNA-alginate amphiphilic binder for silicon and silicon-graphite blended electrodes is reported. Mimicking mucin's structure comprised of a hydrophobic protein backbone and hydrophilic oligosaccharide branches, the renatured DNA-alginate binder offers amphiphilicity from both components, along with a 3D fractal network structure. The DNA-alginate binder facilitates homogeneous distribution of electrode components in the electrode as well as its enhanced adhesion onto a current collector, leading to improved cyclability in both silicon and silicon-graphite blended electrodes. © 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

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.

Automated eddy current analysis of materials

NASA Technical Reports Server (NTRS)

Workman, Gary L.

1991-01-01

The use of eddy current techniques for characterizing flaws in graphite-based filament-wound cylindrical structures is described. A major emphasis was also placed upon incorporating artificial intelligence techniques into the signal analysis portion of the inspection process. Developing an eddy current scanning system using a commercial robot for inspecting graphite structures (and others) was a goal in the overall concept and is essential for the final implementation for the expert systems interpretation. Manual scans, as performed in the preliminary work here, do not provide sufficiently reproducible eddy current signatures to be easily built into a real time expert system. The expert systems approach to eddy current signal analysis requires that a suitable knowledge base exist in which correct decisions as to the nature of a flaw can be performed. A robotic workcell using eddy current transducers for the inspection of carbon filament materials with improved sensitivity was developed. Improved coupling efficiencies achieved with the E-probes and horseshoe probes are exceptional for graphite fibers. The eddy current supervisory system and expert system was partially developed on a MacIvory system. Continued utilization of finite element models for predetermining eddy current signals was shown to be useful in this work, both for understanding how electromagnetic fields interact with graphite fibers, and also for use in determining how to develop the knowledge base. Sufficient data was taken to indicate that the E-probe and the horseshoe probe can be useful eddy current transducers for inspecting graphite fiber components. The lacking component at this time is a large enough probe to have sensitivity in both the far and near field of a thick graphite epoxy component.

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.

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

Experimental high temperature carbon isotope fractionation involving graphite

NASA Astrophysics Data System (ADS)

Kueter, N.; Schmidt, M. W.; Lilley, M. D.; Bernasconi, S. M.

2016-12-01

Graphite/carbonate carbon isotope fractionation was mainly investigated at 400- 800°C and is based on empirical calibrations, theoretical calculations and few experiments [1,2]. Own work on COH-fluid/graphite isotope fractionation shows that in experiments up to 1000oC a fluid phase is always enriched in 13C compared to coexisting graphitic carbon. The eventual kinetic isotope effect in these experiments is best displayed by the graphitic carbon being at least 3 ‰ lighter than methane. Only few experiments done in the graphite/carbonate pair dealt with higher temperatures reaching 1400°C, indicating a fractionation of up to 2 ‰ at temperatures of the Earth's mantle [2-4]. To better understand carbon isotope fractionation in crustal systems and still overcome kinetic effects, we study the graphite/carbonatite pair with piston cylinder experiments in the Na2CO3-CaCO3-CaO-COH system. Tartaric acid (C4H6O6) supplies reduced carbon, time series are performed at 10 kbar, 1300-1800°C. Initial experiments at 1300°C produce well-ordered, micron-sized graphite flakes growing attached to the capsule walls while the Na-Ca-carbonatite-melt quenches to dendritic textures. No gaseous phase was observed. Conditions well above the liquidus of the Na2CO3-CaCO3-binary lead to dissolution of the H2O from tartaric acid decomposition in the melt, any CO2-component is bound by the excess CaO to CaCO3melt while in the relatively oxidizing capsule environment any CH4-component reacts with CO2 to carbon and H2O. The graphite and the carbonatite quench are measured for their δ13C composition using a GasBench II (carbonate-dissolution in phosphoric acid) and TC/EA (residual graphite combusted in oxygen atmosphere) system coupled to a Thermo Fischer IRMS. Our results expand from the graphite-carbonate system to graphite-fluid system when adding available fluid-carbonate fractionation factors, but are also directly applicable to diamond synthesis as graphite is often found as a precursor phase in diamond-growth experiments in carbonatite systems and natural diamonds. [1] Chacko et al. (2001) Rev Min Geochem; Deines & Eggler (2009) GCA; [3] Scheele & Hoefs (1992) CMP; [4] Chacko et al. (1991) GCA

Develop and demonstrate manufacturing processes for fabricating graphite filament reinforced polymide (Gr/PI) composite structural elements

NASA Technical Reports Server (NTRS)

Chase, V. A.; Harrison, E. S.

1985-01-01

A study was conducted to assess the merits of using graphite/polyimide, NR-150B2 resin, for structural applications on advanced space launch vehicles. The program was divided into two phases: (1) Fabrication Process Development; and (2) Demonstration Components. The first phase of the program involved the selection of a graphite fiber, quality assurance of the NR-150B2 polyimide resin, and the quality assurance of the graphite/polyimide prepreg. In the second phase of the program, a limited number of components were fabricated before the NR-150B2 resin system was removed from the market by the supplier, Du Pont. The advancement of the NR-150B2 polyimide resin binder was found to vary significantly based on previous time and temperature history during the prepregging operation. Strength retention at 316C (600F) was found to be 50% that of room temperature strength. However, the composite would retain its initial strength after 200 hours exposure at 316C (600F). Basic chemistry studies are required for determining NR-150B2 resin binder quality assurance parameters. Graphite fibers are available that can withstand high temperature cure and postcure cycles.

Space structures concepts and materials

NASA Technical Reports Server (NTRS)

Nowitzky, A. M.; Supan, E. C.

1988-01-01

An extension is preseted of the evaluation of graphite/aluminum metal matrix composites (MMC) for space structures application. A tubular DWG graphite/aluminum truss assembly was fabricated having the structural integrity and thermal stability needed for space application. DWG is a proprietary thin ply continuous graphite reinforced aluminum composite. The truss end fittings were constructed using the discontinuous ceramic particulate reinforced MMC DWAl 20 (trademark). Thermal stability was incorporated in the truss by utilizing high stiffness, negative coefficient of thermal expansion (CTE) P100 graphite fibers in a 6061 aluminum matrix, crossplied to provide minimized CTE in the assembled truss. Tube CTE was designed to be slightly negative to offset the effects of the end fitting and sleeve, CTE values of which are approx. 1/2 that of aluminum. In the design of the truss configuration, the CTE contribution of each component was evaluated to establish the component dimension and layup configuration required to provide a net zero CTE in the subassemblies which would then translate to a zero CTE for the entire truss bay produced.

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)

The preliminary design of bearings for the control system of a high-temperature lithium-cooled nuclear reactor

NASA Technical Reports Server (NTRS)

Yacobucci, H. G.; Waldron, W. D.; Walowit, J. A.

1973-01-01

The design of bearings for the control system of a fast reactor concept is presented. The bearings are required to operate at temperatures up to 2200 F in one of two fluids, lithium or argon. Basic bearing types are the same regardless of the fluid. Crowned cylindrical journals were selected for radially loaded bearings and modified spherical bearings were selected for bearings under combined thrust and radial loads. Graphite and aluminum oxide are the materials selected for the argon atmosphere bearings while cermet compositions (carbides or nitrides bonded with refractory metals) were selected for the lithium lubricated bearings. Mounting of components is by shrink fit or by axial clamping utilizing differential thermal expansion.

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.

Electrocatalytic N-Doped Graphitic Nanofiber - Metal/Metal Oxide Nanoparticle Composites.

PubMed

Tang, Hongjie; Chen, Wei; Wang, Jiangyan; Dugger, Thomas; Cruz, Luz; Kisailus, David

2018-03-01

Carbon-based nanocomposites have shown promising results in replacing commercial Pt/C as high-performance, low cost, nonprecious metal-based oxygen reduction reaction (ORR) catalysts. Developing unique nanostructures of active components (e.g., metal oxides) and carbon materials is essential for their application in next generation electrode materials for fuel cells and metal-air batteries. Herein, a general approach for the production of 1D porous nitrogen-doped graphitic carbon fibers embedded with active ORR components, (M/MO x , i.e., metal or metal oxide nanoparticles) using a facile two-step electrospinning and annealing process is reported. Metal nanoparticles/nanoclusters nucleate within the polymer nanofibers and subsequently catalyze graphitization of the surrounding polymer matrix and following oxidation, create an interconnected graphite-metal oxide framework with large pore channels, considerable active sites, and high specific surface area. The metal/metal oxide@N-doped graphitic carbon fibers, especially Co 3 O 4 , exhibit comparable ORR catalytic activity but superior stability and methanol tolerance versus Pt in alkaline solutions, which can be ascribed to the synergistic chemical coupling effects between Co 3 O 4 and robust 1D porous structures composed of interconnected N-doped graphitic nanocarbon rings. This finding provides a novel insight into the design of functional electrocatalysts using electrospun carbon nanomaterials for their application in energy storage and conversion fields. © 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

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.

CMB-13 research on carbon and graphite

NASA Technical Reports Server (NTRS)

Smith, M. C.

1972-01-01

The effects of grinding on Santa Maria coke are considered, as well as the production of resin-bonded graphite from the coke. Kynol fibers, properties and purities of coal tar pitches, carbonization of resin components, synthesis of gamma BL (4-furfuryl 2-pentenoic acid gamma lactone), and a glass-like carbon powder for use as a filler are also discussed. The hydrogen contents of commercial cokes and graphites are tabulated, and a quantimet image-analyzing computer and its operation are described.

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

Mohanty, Subhasish; Majumdar, Saurindranath

Irradiation creep plays a major role in the structural integrity of the graphite components in high temperature gas cooled reactors. Finite element procedures combined with a suitable irradiation creep model can be used to simulate the time-integrated structural integrity of complex shapes, such as the reactor core graphite reflector and fuel bricks. In the present work a comparative study was undertaken to understand the effect of linear and nonlinear irradiation creep on results of finite element based stress analysis. Numerical results were generated through finite element simulations of a typical graphite reflector.

Pyrolytic graphite collector development program

NASA Technical Reports Server (NTRS)

Wilkins, W. J.

1982-01-01

Pyrolytic graphite promises to have significant advantages as a material for multistage depressed collector electrodes. Among these advantages are lighter weight, improved mechanical stiffness under shock and vibration, reduced secondary electron back-streaming for higher efficiency, and reduced outgassing at higher operating temperatures. The essential properties of pyrolytic graphite and the necessary design criteria are discussed. This includes the study of suitable electrode geometries and methods of attachment to other metal and ceramic collector components consistent with typical electrical, thermal, and mechanical requirements.

Low cost damage tolerant composite fabrication

NASA Technical Reports Server (NTRS)

Palmer, R. J.; Freeman, W. T.

1988-01-01

The resin transfer molding (RTM) process applied to composite aircraft parts offers the potential for using low cost resin systems with dry graphite fabrics that can be significantly less expensive than prepreg tape fabricated components. Stitched graphite fabric composites have demonstrated compression after impact failure performance that equals or exceeds that of thermoplastic or tough thermoset matrix composites. This paper reviews methods developed to fabricate complex shape composite parts using stitched graphite fabrics to increase damage tolerance with RTM processes to reduce fabrication cost.

Method for depositing a uniform layer of particulate material on the surface of an article having interconnected porosity

DOEpatents

Wrenn, Jr., George E.; Lewis, Jr., John

1984-01-01

The invention is a method for depositing liquid-suspended particles on an immersed porous article characterized by interconnected porosity. In one form of the invention, coating is conducted in a vessel containing an organic liquid supporting a colloidal dispersion of graphite sized to lodge in surface pores of the article. The liquid comprises a first volatile component (e.g., acetone) and a second less-volatile component (e.g., toluene) containing a dissolved organic graphite-bonding agent. The liquid also contains an organic agent (e.g., cellulose gum) for maintaining the particles in suspension. A porous carbon article to be coated is immersed in the liquid so that it is permeated therewith. While the liquid is stirred to maintain a uniform blend, the vessel headspace is evacuated to effect flashing-off of the first component from the interior of the article. This causes particle-laden liquid exterior of the article to flow inwardly through its surface pores, lodging particles in these pores and forming a continuous graphite coating. The coated article is retrieved and heated to resin-bond the graphite. The method can be used to form a smooth, adherent, continuous coating of various materials on various porous articles. The method is rapid and reproducible.

Method for depositing a uniform layer of particulate material on the surface of an article having interconnected porosity

DOEpatents

Wrenn, G.E. Jr.; Lewis, J. Jr.

1982-09-29

The invention is a method for depositing liquid-suspended particles on an immersed porous article characterized by interconnected porosity. In one form of the invention, coating is conducted in a vessel containing an organic liquid supporting a colloidal dispersion of graphite sized to lodge in surface pores of the article. The liquid comprises a first volatile component (e.g., acetone) and a second less-volatile component (e.g., toluene) containing a dissolved organic graphite-bonding agent. The liquid also contains an organic agent (e.g., cellulose gum) for maintaining the particles in suspension. A porous carbon article to be coated is immersed in the liquid so that it is permeated therewith. While the liquid is stirred to maintain a uniform blend, the vessel headspace is evacuated to effect flashing-off of the first component from the interior of the article. This causes particle-laden liquid exterior of the article to flow inwardly through its surface pores, lodging particles in these pores and forming a continuous graphite coating. The coated article is retrieved and heated to resin-bond the graphite. The method can be used to form a smooth, adherent, continuous coating of various materials on various porous articles. The method is rapid and reproducible.

Adhesive properties and adhesive joints strength of graphite/epoxy composites

NASA Astrophysics Data System (ADS)

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

2017-05-01

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

Binary gaseous mixture and single component adsorption of methane and argon on exfoliated graphite

NASA Astrophysics Data System (ADS)

Russell, Brice Adam

Exfoliated graphite was used as a substrate for adsorption of argon and methane. Adsorption experiments were conducted for both equal parts mixtures of argon and methane and for each gas species independently. The purpose of this was to compare mixture adsorption to single component adsorption and to investigate theoretical predictions concerning the kinetics of adsorption made by Burde and Calbi.6 In particular, time to reach pressure equilibrium of a single dose at a constant temperature for the equal parts mixture was compared to time of adsorption for each species by itself. It was shown that mixture adsorption is a much more complex and time consuming process than single component adsorption and requires a much longer amount of time to reach equilibrium. Information about the composition evolution of the mixture during the times when pressure was going toward equilibrium was obtained using a quadrupole mass spectrometer. Evidence for initial higher rate of adsorption for the weaker binding energy species (argon) was found as well as overall composition change which clearly indicated a higher coverage of methane on the graphite sample by the time equilibration was reached. Effective specific surface area of graphite for both argon and methane was also determined using the Point-B method.2

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...

Program For Optimization Of Nuclear Rocket Engines

NASA Technical Reports Server (NTRS)

Plebuch, R. K.; Mcdougall, J. K.; Ridolphi, F.; Walton, James T.

1994-01-01

NOP is versatile digital-computer program devoloped for parametric analysis of beryllium-reflected, graphite-moderated nuclear rocket engines. Facilitates analysis of performance of engine with respect to such considerations as specific impulse, engine power, type of engine cycle, and engine-design constraints arising from complications of fuel loading and internal gradients of temperature. Predicts minimum weight for specified performance.

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.

Baking and helium glow discharge cleaning of SST-1 Tokamak with graphite plasma facing components

NASA Astrophysics Data System (ADS)

Semwal, P.; Khan, Z.; Raval, D. C.; Dhanani, K. R.; George, S.; Paravastu, Y.; Prakash, A.; Thankey, P.; Ramesh, G.; Khan, M. S.; Saikia, P.; Pradhan, S.

2017-04-01

Graphite plasma facing components (PFCs) were installed inside the SST-1 vacuum vessel. Prior to installation, all the graphite tiles were baked at 1000 °C in a vacuum furnace operated below 1.0 × 10-5 mbar. However due to the porous structure of graphite, they absorb a significant amount of water vapour from air during the installation process. Rapid desorption of this water vapour requires high temperature bake-out of the PFCs at ≥ 250 °C. In SST-1 the PFCs were baked at 250 °C using hot nitrogen gas facility to remove the absorbed water vapour. Also device with large graphite surface area has the disadvantage that a large quantity of hydrogen gets trapped inside it during plasma discharges which makes density control difficult. Helium glow discharge cleaning (He-GDC) effectively removes this stored hydrogen as well as other impurities like oxygen and hydrocarbon within few nano-meters from the surface by particle induced desorption. Before plasma operation in SST-1 tokamak, both baking of PFCs and He-GDC were carried out so that these impurities were removed effectively. The mean desorption yield of hydrogen was found to be 0.24. In this paper the results of baking and He-GDC experiments of SST-1 will be presented in detail.

All about the Pencil

ERIC Educational Resources Information Center

Stevens, Sue

2006-01-01

This article presents the history of pencils. Graphite, which is the main component that makes pencils write, was first discovered to be useful in marking the sheep of local farmers in Borrowdale, England in 1954. This graphite left a much darker mark than lead, which made it ideal for use by writers and artists. Around, 1795, Nicolas-Jacques…

Graphitic biocarbon from metal-catalyzed hydrothermal carbonization of lignin

DOE PAGES

Demir, Muslum; Kahveci, Zafer; Aksoy, Burak; ...

2015-10-09

Lignin is a high-volume byproduct from the pulp and paper industry and is currently burned to generate electricity and process heat. Moreover, the industry has been searching for high value-added uses of lignin to improve the process economics. In addition, battery manufacturers are seeking nonfossil sources of graphitic carbon for environmental sustainability. In our work, lignin (which is a cross-linked polymer of phenols, a component of biomass) is converted into graphitic porous carbon using a two-step conversion. Lignin is first carbonized in water at 300 °C and 1500 psi to produce biochar, which is then graphitized using a metal nitratemore » catalyst at 900–1100 °C in an inert gas at 15 psi. Graphitization effectiveness of three different catalysts—iron, cobalt, and manganese nitrates—is examined. The product is analyzed for morphology, thermal stability, surface properties, and electrical conductivity. Both temperature and catalyst type influenced the degree of graphitization. A good quality graphitic carbon was obtained using catalysis by Mn(NO 3) 2 at 900 °C and Co(NO 3) 2 at 1100 °C.« less

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.

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

Recompressed exfoliated graphite articles

DOEpatents

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

2013-08-06

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

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

Ultrafast plasmon-enhanced hot electron process in model heterojunctions: Ag/TiO2 and Ag/graphite

NASA Astrophysics Data System (ADS)

Petek, Hrvoje

We study the plasmonically enhanced nonlinear photoemission from Ag nanocluster-decorated graphite and TiO2(110) surfaces by time-resolved two-photon photoemission spectroscopy (TR-2PP). Evaporating Ag atoms on graphite and TiO2 surfaces forms pancake-like Ag clusters with 5 nm diameter and 1-1.5 nm height through self-limiting growth mode. The Ag nanoparticles enhance the two-photon photoemission (2PP) signal by approximately two-orders of magnitude as compared with the bare surfaces for p-polarized excitation. In the case of s-polarization there is essentially no enhancement for graphite, and only about an order-of-magnitude enhancement for TiO2. Wavelength dependent measurements of the enhancement reveal that for Ag/graphite there is a single plasmonic resonance due to the ⊥-plasmon mode at 3.6 eV. By contrast, for Ag/TiO2 there are ⊥ and ||-plasmon modes with resonant energies of 3.8 and 3.1 eV, respectively. Apparently the dielectric properties of the substrate have strong influence on the type and frequency of Ag plasmonic modes that can exist on the surfaces. 2PP spectra of the Ag/graphite and Ag/TiO2 surfaces reveal two distinct components that are common to both. The high energy component consists of a coherent 2PP process from an occupied interface state, which only exists in the presence of Ag. We identify this state, as an interface state formed by charge donation from the Ag-5s band to the unoccupied states of the substrates. The low energy component consists of a hot electron signal that is created by plasmon dephasing. TR-2PP measurements are performed on the plasmon-induced electron dynamics to assess their relevance for plasmonically enhanced femtochemistry. This research was supported by NSF Grant CHE-1414466.

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

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

Corrosion of graphite composites in phosphoric acid fuel cells

NASA Technical Reports Server (NTRS)

Christner, L. G.; Dhar, H. P.; Farooque, M.; Kush, A. K.

1986-01-01

Polymers, polymer-graphite composites and different carbon materials are being considered for many of the fuel cell stack components. Exposure to concentrated phosphoric acid in the fuel cell environment and to high anodic potential results in corrosion. Relative corrosion rates of these materials, failure modes, plausible mechanisms of corrosion and methods for improvement of these materials are investigated.

Analysis and measurements of low frequency lightning component penetration through aerospace vehicle metal and graphite skins

NASA Technical Reports Server (NTRS)

Robb, J. D.; Chen, T.

1980-01-01

An analysis of the shielding properties of mixed metal and graphite composite structures has illustrated some important aspects of electromagnetic field penetration into the interior. These include: (1) that graphite access doors on metallic structures will attenuate lightning magnetic fields very little; conversely, metal doors on a graphite structure will also attenuate fields from lightning strike currents very little, i.e., homogeneity of the shield is a critical factor in shielding and (2) that continuous conductors between two points inside a graphite skin such as an air data probe metallic tubing connection to an air data computer can allow large current penetrations into a vehicle interior. The true weight savings resulting from the use of composite materials can only be evaluated after the resulting electromagnetic problems such as current penetrations have been solved, and this generally requires weight addition in the form of cable shields, conductor bonding or external metallization.

On the Application of Lithium Additives in the Electrolytic Production of Primary Aluminum

NASA Astrophysics Data System (ADS)

Saitov, A. V.; Bazhin, V. Yu.; Povarov, V. G.

2017-12-01

The behavior of carbon-graphite subjected to treatment in the lithium carbonate Li2CO3 melt without cryolite and alkali-metal fluorides is studied to reliably estimate the influence of lithium on the surface layers of a carbon-containing cathode lining. The chemical composition and the structure of the carbon-graphite material after its interaction with lithium in the Li2CO3 melt have been studied. The high-temperature interaction of the system components in the melt is found to be accompanied by fracture of the operating surface of the carbon-graphite material, while the carbon-graphite surface does not failed upon interacting with lithium vapors. Based on the obtained data, a model for the formation of lithium ions during the reduction of lithium and its interaction with a carbon-graphite sample during the electrolysis of lithium carbonate is proposed.

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)

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

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

INVESTIGATION OF TITANIUM BONDED GRAPHITE FOAM COMPOSITES FOR MICRO ELECTRONIC MECHANICAL SYSTEMS (MEMS) APPLICATIONS

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

Menchhofer, Paul A.

PiMEMS Inc. (Santa Barbara, CA) in collaboration with ORNL investigated the use of Titanium Bonded Graphite Foam Composites (TBGC) for thermal mitigation in Micro Electronic Mechanical Systems (MEMS) applications. Also considered were potentially new additive manufacturing routes to producing novel high surface area micro features and diverse shaped heat transfer components for numerous lightweight MEMs applications.

Gelcasting polymeric precursors for producing net-shaped graphites

DOEpatents

Klett, James W.; Janney, Mark A.

2002-01-01

The present invention discloses a method for molding complex and intricately shaped high density monolithic carbon, carbon-carbon, graphite, and thermoplastic composites using gelcasting technology. The method comprising a polymeric carbon precursor, a solvent, a dispersant, an anti-foaming agent, a monomer system, and an initiator system. The components are combined to form a suspension which is poured into a mold and heat-treated to form a thermoplastic part. The thermoplastic part can then be further densified and heat-treated to produce a high density carbon or graphite composite. The present invention also discloses the products derived from this method.

B{sub 4}C-SiC reaction-sintered coatings on graphite plasma facing components

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

Valentine, P.G.; Trester, P.W.; Winter, J.

1994-05-01

Boron carbide plus silicon carbide (B{sub 4}C-SiC) reaction-sintered coatings for use on graphite plasma-facing components were developed. Such coatings are of interest in TEXTOR tokamak limiter-plasma interactions as a means of reducing carbon erosion, of providing a preferred release of boron for oxygen gettering, and of investigating silicon`s effect on radiative edge phenomena. Specimens evaluated had (a) either Ringsdorfwerke EK 98 graphite or Le Carbon Lorraine felt-type AEROLOR A05 CFC substrates; (b) multiphase coatings, comprised of B{sub 4}C, Sic, and graphite; (c) nominal coating compositions of 69 wt.-% B{sub 4}C + 31 wt.-% SiC; and (d) nominal coating thicknesses betweenmore » 250 and 775 {mu}m. Coated coupons were evaluated by high heat flux experiments in the JUDITH (electron beam) test facility at KFA. Simulated disruptions, with energy densities up to 10 MJm{sup {minus}2}, and normal operation simulations, with power densities up to 12 MWm{sup {minus}2}, were conducted. The coatings remained adherent; at the highest levels tested, minor changes occurred, including localized remelting, modification of the crystallographic phases, occasional microcracking, and erosion.« less

Comparison of irradiation behaviour of HTR graphite grades

NASA Astrophysics Data System (ADS)

Heijna, M. C. R.; de Groot, S.; Vreeling, J. A.

2017-08-01

The INNOGRAPH irradiations were executed in the High Flux Reactor (HFR) in Petten by NRG supported by the European Framework programs HTR-M, RAPHAEL, and ARCHER to generate data on the irradiation behaviour of graphite grades for High Temperature Reactor (HTR) application available at that time. Samples of the graphite grades NBG-10, NBG-17, NBG-18, NBG-20, NBG-25, PCEA, PPEA, PCIB, and IG-110 have been irradiated at 750 °C and 950 °C. The inherent scatter induced by the probabilistic material behaviour of graphite requires uncertainty and scatter induced by test conditions and post-irradiation examination to be minimized. The INNOGRAPH irradiations supplied an adequate number of irradiated samples to enable accurate determination of material properties and their evolution under irradiation. This allows comparison of different graphite grades and a qualitative assessment of their appropriateness for HTR applications, as a basis of selection, design and core component lifetime. The results indicate that coarse grained graphite grades exhibit more favourable behaviour for application in HTRs due to their low dimensional anisotropy and fracture propagation resilience.

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.

A Low O/Si Ratio on the Surface of Mercury: Evidence for Silicon Smelting?

NASA Astrophysics Data System (ADS)

McCubbin, Francis M.; Vander Kaaden, Kathleen E.; Peplowski, Patrick N.; Bell, Aaron S.; Nittler, Larry R.; Boyce, Jeremy W.; Evans, Larry G.; Keller, Lindsay P.; Elardo, Stephen M.; McCoy, Timothy J.

2017-10-01

Data from the Gamma-Ray Spectrometer (GRS) that flew on the MErcury Surface, Space ENvironment, GEochemistry, and Ranging spacecraft indicate that the O/Si weight ratio of Mercury's surface is 1.2 ± 0.1. This value is lower than any other celestial surface that has been measured by GRS and suggests that 12-20% of the surface materials on Mercury are composed of Si-rich, Si-Fe alloys. The origin of the metal is best explained by a combination of space weathering and graphite-induced smelting. The smelting process would have been facilitated by interaction of graphite with boninitic and komatiitic parental liquids. Graphite entrained at depth would have reacted with FeO components dissolved in silicate melt, resulting in the production of up to 0.4-0.9 wt % CO from the reduction of FeO to Fe0—CO production that could have facilitated explosive volcanic processes on Mercury. Once the graphite-entrained magmas erupted, the tenuous atmosphere on Mercury prevented the buildup of CO over the lavas. The partial pressure of CO would have been sufficiently low to facilitate reaction between graphite and SiO2 components in silicate melts to produce CO and metallic Si. Although exotic, Si-rich metal as a primary smelting product is hypothesized on Mercury for three primary reasons: (1) low FeO abundances of parental magmas, (2) elevated abundances of graphite in the crust and regolith, and (3) the presence of only a tenuous atmosphere at the surface of the planet within the 3.5-4.1 Ga timespan over which the planet was resurfaced through volcanic processes.

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.

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.

High temperature resin matrix composites for aerospace structures

NASA Technical Reports Server (NTRS)

Davis, J. G., Jr.

1980-01-01

Accomplishments and the outlook for graphite-polyimide composite structures are briefly outlined. Laminates, skin-stiffened and honeycomb sandwich panels, chopped fiber moldings, and structural components were fabricated with Celion/LARC-160 and Celion/PMR-15 composite materials. Interlaminar shear and flexure strength data obtained on as-fabricated specimens and specimens that were exposed for 125 hours at 589 K indicate that epoxy sized and polyimide sized Celion graphite fibers exhibit essentially the same behavior in a PMR-15 matrix composite. Analyses and tests of graphite-polyimide compression and shear panels indicate that utilization in moderately loaded applications offers the potential for achieving a 30 to 50 percent reduction in structural mass compared to conventional aluminum panels. Data on effects of moisture, temperature, thermal cycling, and shuttle fluids on mechanical properties indicate that both LARC-160 and PMR-15 are suitable matrix materials for a graphite-polyimide aft body flap. No technical road blocks to building a graphite-polyimide composite aft body flap are identified.

Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

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

Jason M. Harp; Paul A. Demkowicz

2014-10-01

In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10 -4 to 10 -5) of as manufactured and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materialsmore » is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application is considered. To test the feasibility, the response of the PGS system to idealized fission product distributions has been simulated using Monte Carlo radiation transport simulations. Previous work that applied similar techniques during the AGR-1 experiment will also be discussed as well as planned uses for the GECT technique during the post irradiation examination of the AGR-2 experiment. The GECT technique has also been applied to other irradiated nuclear fuel systems that were currently available in the HFEF hot cell including oxide fuel pins, metallic fuel pins, and monolithic plate fuel.« less

Processing, Microstructure, and Mechanical Properties of Interpenetrating Biomorphic Graphite/Copper Composites

NASA Astrophysics Data System (ADS)

Childers, Amanda Esther Sall

Composite properties can surpass those of the individual phases, allowing for the development of advanced, high-performance materials. Bio-inspired and naturally-derived materials have garnered attention as composite constituents due to their inherently efficient and complex structures. Wood-derived ceramics, produced by converting a wood precursor into a ceramic scaffold, can exhibit a wide range of microstructures depending on the wood species, including porosity, pore size and distribution, and connectivity. The focus of this work was to investigate the processing, microstructure, and properties of graphite/copper composites produced using wood-derived graphite scaffolds. Graphite/copper composites combine low specific gravity, high thermal conductivity, and tailorable thermal expansion properties, and due to the non-wetting behavior of copper to graphite, offer a unique system in which mechanically bonded interfaces in composites can be studied. Graphite scaffolds were produced from red oak, beech, and pine precursors using a catalytic pyrolyzation method, resulting in varying types of pore networks. Two infiltration methods were investigated to overcome challenges associated with non-wetting systems: copper electrodeposition and pressure-assisted melt infiltration. The phase distributions, constituent properties, interfacial characteristics, mechanical behavior, and load partitioning of these biomorphic graphite/copper composites were investigated, and were correlated to the wood species. The multi-domain feature sizes in the graphite scaffolds resulted in composites with copper relegated not only to the large, connected channels produced from the transport features in the wood, but also within the smaller, lower aspect ratio fibrous regions of the scaffold. Both features contributed to the mechanical behavior of the composites to varying degrees depending on the wood species. A multi-component predictive model also was developed and used to guide the additive-assisted electroplating of the graphitized scaffold, and helped illuminate the roles of plating additives in macro-sized channels. The model can be adapted for many material systems, sample geometries, and plating conditions to investigate the use of metal electrodeposition as a means of scaffold infiltration. Additionally, X-ray diffraction tomography was used to resolve position-dependent strain in a composite. The results of this nascent capability were discussed with respect to a two-component system under increasing uniaxial load, and compared to the results of conventional volume-averaged measurements.

High-temperature, high-pressure bonding of nested tubular metallic components

DOEpatents

Quinby, T.C.

A tool is described for effecting high-temperature, high-compression bonding between the confronting faces of nested, tubular, metallic components. In a typical application, the tool is used to produce tubular target assemblies for irradiation in nuclear reactors or particle accelerators. The target assembly comprising a uranum foil and an aluninum-alloy substrate. The tool is composed of graphite. It comprises a tubular restraining member in which a mechanically expandable tubular core is mounted to form an annulus. The components to be bonded are mounted in nested relation in the annulus. The expandable core is formed of individually movable, axially elongated segments whose outer faces cooperatively define a cylindrical pressing surface and whose inner faces cooperatively define two opposed, inwardly tapered, axial bores. Tapered rams extend into the bores. The loaded tool is mounted in a conventional hot-press provided with evacuation means, heaters for maintaining its interior at bonding temperature, and hydraulic cylinders for maintaining a selected inwardly directed pressure on the tapered rams. With the hot-press evacuated and the loaded tool at the desired temperature, the cylinders are actuated to apply the selected pressure to the rams. The rams in turn expand the segmented core to maintain the nested components in compression against the restraining member. These conditions are maintained until the confronting faces of the nested components are joined in a continuous, uniform bond characterized by high thermal conductivity.

The Investigation and Development of Low Cost Hardware Components for Proton-Exchange Membrane Fuel Cells - Final Report

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

George A. Marchetti

1999-12-15

Proton exchange membrane (PEM) fuel cell components, which would have a low-cost structure in mass production, were fabricated and tested. A fuel cell electrode structure, comprising a thin layer of graphite (50 microns) and a front-loaded platinum catalyst layer (600 angstroms), was shown to produce significant power densities. In addition, a PEM bipolar plate, comprising flexible graphite, carbon cloth flow-fields and an integrated polymer gasket, was fabricated. Power densities of a two-cell unit using this inexpensive bipolar plate architecture were shown to be comparable to state-of-the-art bipolar plates.

Eddy current inspection of graphite fiber components

NASA Technical Reports Server (NTRS)

Workman, G. L.; Bryson, C. C.

1990-01-01

The recognition of defects in materials properties still presents a number of problems for nondestructive testing in aerospace systems. This project attempts to utilize current capabilities in eddy current instrumentation, artificial intelligence, and robotics in order to provide insight into defining geometrical aspects of flaws in composite materials which are capable of being evaluated using eddy current inspection techniques. The unique capabilities of E-probes and horseshoe probes for inspecting probes for inspecting graphite fiber materials were evaluated and appear to hold great promise once the technology development matures. The initial results are described of modeling eddy current interactions with certain flaws in graphite fiber samples.

Graphite/epoxy composite adapters for the Space Shuttle/Centaur vehicle

NASA Technical Reports Server (NTRS)

Kasper, Harold J.; Ring, Darryl S.

1990-01-01

The decision to launch various NASA satellite and Air Force spacecraft from the Space Shuttle created the need for a high-energy upper stage capable of being deployed from the cargo bay. Two redesigned versions of the Centaur vehicle which employed a graphite/epoxy composite material for the forward and aft adapters were selected. Since this was the first time a graphite/epoxy material was used for Centaur major structural components, the development of the adapters was a major effort. An overview of the composite adapter designs, subcomponent design evaluation test results, and composite adapter test results from a full-scale vehicle structural test is presented.

Nature of very small grains - PAH molecules or silicates?. [Polycyclic Aromatic Hydrocarbon in interstellar dust

NASA Technical Reports Server (NTRS)

Desert, F. X.; Leger, A.; Puget, J. L.; Boulanger, F.; Sellgren, K.

1986-01-01

The predictions of the model of Puget et al. (1985) for the emission from Very Small Grains (VSGs) including both graphitic and silicate components are compared with published 8-13-micron observations of astronomical sources. The VSGs are found to be mainly graphitic and an upper limit is placed on the relative mass of silicates based on lack of the 9.7-micron silicate emission feature on M 82 and NGC 2023. This dissymetry in the composition of VSGs supports the suggestion that they are formed in grain-grain collisions where the behaviors of graphite and silicate grains are expected to be quite different.

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.

Compacted graphite iron: Cast iron makes a comeback

NASA Astrophysics Data System (ADS)

Dawson, S.

1994-08-01

Although compacted graphite iron has been known for more than four decades, the absence of a reliable mass-production technique has resulted in relatively little effort to exploit its operational benefits. However, a proven on-line process control technology developed by SinterCast allows for series production of complex components in high-quality CGI. The improved mechanical properties of compacted graphite iron relative to conventional gray iron allow for substantial weight reduction in gasoline and diesel engines or substantial increases in horsepower, or an optimal combination of both. Concurrent with these primary benefits, CGI also provides significant emissions and fuel efficiency benefits allowing automakers to meet legislated performance standards. The operational and environmental benefits of compacted graphite iron together with its low cost and recyclability reinforce cast iron as a prime engineering material for the future.

A near infra-red video system as a protective diagnostic for electron cyclotron resonance heating operation in the Wendelstein 7-X stellarator

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

Preynas, M.; Laqua, H. P.; Marsen, S.

The Wendelstein 7-X stellarator is a large nuclear fusion device based at Max-Planck-Institut für Plasmaphysik in Greifswald in Germany. The main plasma heating system for steady state operation in W7-X is electron cyclotron resonance heating (ECRH). During operation, part of plama facing components will be directly heated by the non-absorbed power of 1 MW rf beams of ECRH. In order to avoid damages of such components made of graphite tiles during the first operational phase, a near infra-red video system has been developed as a protective diagnostic for safe and secure ECRH operation. Both the mechanical design housing the cameramore » and the optical system are very flexible and respect the requirements of steady state operation. The full system including data acquisition and control system has been successfully tested in the vacuum vessel, including on-line visualization and data storage of the four cameras equipping the ECRH equatorial launchers of W7-X.« less

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.

Advanced Materials and Fabrication Techniques for the Orion Attitude Control Motor

NASA Technical Reports Server (NTRS)

Gorti, Sridhar; Holmes, Richard; O'Dell, John; McKechnie, Timothy; Shchetkovskiy, Anatoliy

2013-01-01

Rhenium, with its high melting temperature, excellent elevated temperature properties, and lack of a ductile-to-brittle transition temperature (DBTT), is ideally suited for the hot gas components of the ACM (Attitude Control Motor), and other high-temperature applications. However, the high cost of rhenium makes fabricating these components using conventional fabrication techniques prohibitive. Therefore, near-net-shape forming techniques were investigated for producing cost-effective rhenium and rhenium alloy components for the ACM and other propulsion applications. During this investigation, electrochemical forming (EL-Form ) techniques were evaluated for producing the hot gas components. The investigation focused on demonstrating that EL-Form processing techniques could be used to produce the ACM flow distributor. Once the EL-Form processing techniques were established, a representative rhenium flow distributor was fabricated, and samples were harvested for material properties testing at both room and elevated temperatures. As a lower cost and lighter weight alternative to an all-rhenium component, rhenium- coated graphite and carbon-carbon were also evaluated. The rhenium-coated components were thermal-cycle tested to verify that they could withstand the expected thermal loads during service. High-temperature electroforming is based on electrochemical deposition of compact layers of metals onto a mandrel of the desired shape. Mandrels used for electro-deposition of near-net shaped parts are generally fabricated from high-density graphite. The graphite mandrel is easily machined and does not react with the molten electrolyte. For near-net shape components, the inner surface of the electroformed part replicates the polished graphite mandrel. During processing, the mandrel itself becomes the cathode, and scrap or refined refractory metal is the anode. Refractory metal atoms from the anode material are ionized in the molten electrolytic solution, and are deposited onto the cathodic mandrel by electrochemical reduction. Rotation of the mandrel ensures uniform distribution of refractory material. The EL-Form process allows for manufacturing in an inert atmosphere with deposition rates from 0.0004 to 0.002 in./h (10.2 to 50.8 m/h). Thicknesses typically range from microns to greater than 0.5 in. (13 mm). The refractory component produced is fabricated, dependably, to within one micron of the desired tolerances with no shrinkage or distortion as in other refractory metal manufacture techniques. The electroforming process has been used to produce solid, nonporous deposits of rhenium, iridium, niobium, tungsten, and their alloys.

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

Molecular dynamics investigations of liquid-vapor interaction and adsorption of formaldehyde, oxocarbons, and water in graphitic slit pores.

PubMed

Huang, Pei-Hsing; Hung, Shang-Chao; Huang, Ming-Yueh

2014-08-07

Formaldehyde exposure has been associated with several human cancers, including leukemia and nasopharyngeal carcinoma, motivating the present investigation on the microscopic adsorption behaviors of formaldehyde in multi-component-mixture-filled micropores. Molecular dynamics (MD) simulation was used to investigate the liquid-vapor interaction and adsorption of formaldehyde, oxocarbons, and water in graphitic slit pores. The effects of the slit width, system temperature, concentration, and the constituent ratio of the mixture on the diffusion and adsorption properties are studied. As a result of interactions between the components, the z-directional self-diffusivity (D(z)) in the mixture substantially decreased by about one order of magnitude as compared with that of pure (single-constituent) adsorbates. When the concentration exceeds a certain threshold, the D(z) values dramatically decrease due to over-saturation inducing barriers to diffusion. The binding energy between the adsorbate and graphite at the first adsorption monolayer is calculated to be 3.99, 2.01, 3.49, and 2.67 kcal mol(-1) for CO2, CO, CH2O, and H2O, respectively. These values agree well with those calculated using the density functional theory coupled cluster method and experimental results. A low solubility of CO2 in water and water preferring to react with CH2O, forming hydrated methanediol clusters, are observed. Because the cohesion in a hydrated methanediol cluster is much higher than the adhesion between clusters and the graphitic surface, the hydrated methanediol clusters were hydrophobic, exhibiting a large contact angle on graphite.

Visible and infrared optical properties of stacked cone graphite microtubes.

PubMed

Bruce, Charles W; Alyones, Sharhabeel

2012-06-01

The absorptive and scattering optical properties of heat-treated, vapor-grown, graphite microtubes consisting of nanotubes in a "stacked cone" configuration were investigated through the visible and infrared wavelengths using photoacoustic and other spectrometric techniques. However, computations of these properties involved uncertainties that were not easily resolved; the appropriate dielectric coefficients were presumed to be a combination of the published values for the distinct orientations of graphite, but the correct proportions are not evident and none of the reasonable choices produced satisfactory agreement (within the measurement limits of error). Since both of the primary components of the extinction were measured, the appropriate computational codes were employed in reverse to compute the dielectric coefficients for the graphite microtubes. Differences, primarily for the imaginary index, are most distinct for visible and near infrared wavelengths; in this wavelength region, the imaginary index falls progressively to less than half that for the computed mixture.

An investigation on the effects of phase change material on material components used for high temperature thermal energy storage system

NASA Astrophysics Data System (ADS)

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

2016-05-01

The latent heat thermal energy storage (LHTES) systems for concentrated solar power (CSP) plants with advanced power cycle require high temperature phase change materials (PCMs), Graphite foams with high thermal conductivity to enhance the poor thermal conductivity of PCMs. Brazing of the graphite foams to the structural metals of the LHTES system could be a method to assemble the system and a method to protect the structural metals from the molten salts. In the present study, the LHTES prototype capsules using MgCl2-graphite foam composites were assembled by brazing and welding, and tested to investigate the corrosion attack of the PCM salt on the BNi-4 braze. The microstructural analysis showed that the BNi-4 braze alloy can be used not only for the joining of structure alloy to graphite foams but also for the protecting of structure alloy from the corrosion by PCM.

Thermal Pyrolytic Graphite Enhanced Components

NASA Technical Reports Server (NTRS)

Hardesty, Robert E. (Inventor)

2015-01-01

A thermally conductive composite material, a thermal transfer device made of the material, and a method for making the material are disclosed. Apertures or depressions are formed in aluminum or aluminum alloy. Plugs are formed of thermal pyrolytic graphite. An amount of silicon sufficient for liquid interface diffusion bonding is applied, for example by vapor deposition or use of aluminum silicon alloy foil. The plugs are inserted in the apertures or depressions. Bonding energy is applied, for example by applying pressure and heat using a hot isostatic press. The thermal pyrolytic graphite, aluminum or aluminum alloy and silicon form a eutectic alloy. As a result, the plugs are bonded into the apertures or depressions. The composite material can be machined to produce finished devices such as the thermal transfer device. Thermally conductive planes of the thermal pyrolytic graphite plugs may be aligned in parallel to present a thermal conduction path.

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.

The vulnerability of commercial aircraft avionics to carbon fibers

NASA Technical Reports Server (NTRS)

Meyers, J. A.; Salmirs, S.

1980-01-01

Avionics components commonly used in commercial aircraft were tested for vulnerability to failure when operated in an environment with a high density of graphite fibers. The components were subjected to a series of exposures to graphite fibers of different lengths. Lengths used for the tests were (in order) 1 mm, 3 mm, and 10 mm. The test procedure included subjecting the equipment to characteristic noise and shock environments. Most of the equipment was invulnerable or did not fail until extremely high average exposures were reached. The single exception was an air traffic control transponder produced in the early 1960's. It had the largest case open area through which fibers could enter and it had no coated boards.

Environmentally-friendly oxygen-free roasting/wet magnetic separation technology for in situ recycling cobalt, lithium carbonate and graphite from spent LiCoO2/graphite lithium batteries.

PubMed

Li, Jia; Wang, Guangxu; Xu, Zhenming

2016-01-25

The definite aim of the present paper is to present some novel methods that use oxygen-free roasting and wet magnetic separation to in situ recycle of cobalt, Lithium Carbonate and Graphite from mixed electrode materials. The in situ recycling means to change waste into resources by its own components, which is an idea of "waste+waste→resources." After mechanical scraping the mixed electrode materials enrich powders of LiCoO2 and graphite. The possible reaction between LiCoO2 and graphite was obtained by thermodynamic analysis. The feasibility of the reaction at high temperature was studied with the simultaneous thermogravimetry analysis under standard atmospheric pressure. Then the oxygen-free roasting/wet magnetic separation method was used to transfer the low added value mixed electrode materials to high added value products. The results indicated that, through the serious technologies of oxygen-free roasting and wet magnetic separation, mixture materials consist with LiCoO2 and graphite powders are transferred to the individual products of cobalt, Lithium Carbonate and Graphite. Because there is not any chemical solution added in the process, the cost of treating secondary pollution can be saved. This study provides a theoretical basis for industrial-scale recycling resources from spent LIBs. Copyright © 2015 Elsevier B.V. All rights reserved.

Data Report on Post-Irradiation Dimensional Change of AGC-1 Samples

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

William Windes

This report summarizes the initial dimensional changes for loaded and unloaded AGC-1 samples. The dimensional change for all samples is presented as a function of dose. The data is further presented by graphite type and applied load levels to illustrate the differences between graphite forming processes and stress levels within the graphite components. While the three different loads placed on the samples have been verified [ ref: Larry Hull’s report] verification of the AGC-1 sample temperatures and dose levels are expected in the summer of 2012. Only estimated dose and temperature values for the samples are presented in this reportmore » to allow a partial analysis of the results.« less

Dielectric properties of novel polyurethane-PZT-graphite foam composites

NASA Astrophysics Data System (ADS)

Tolvanen, Jarkko; Hannu, Jari; Nelo, Mikko; Juuti, Jari; Jantunen, Heli

2016-09-01

Flexible foam composite materials offer multiple benefits to future electronic applications as the rapid development of the electronics industry requires smaller, more efficient, and lighter materials to further develop foldable and wearable applications. The aims of this work were to examine the electrical properties of three- and four-phase novel foam composites in different conditions, find the optimal mixture for four-phase foam composites, and study the combined effects of lead zirconate titanate (PZT) and graphite fillers. The flexible and highly compressible foams were prepared in a room-temperature mixing process using polyurethane, PZT, and graphite components as well as their combinations, in which air acted as one phase. In three-phase foams the amount of PZT varied between 20 and 80 wt% and the amount of graphite, between 1 and 15 wt%. The four-phase foams were formed by adding 40 wt% of PZT while the amount of graphite ranged between 1 and 15 wt%. The presented results and materials could be utilized to develop new flexible and soft sensor applications by means of material technology.

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

Synthesis of monolithic graphene – graphite integrated electronics

PubMed Central

Park, Jang-Ung; Nam, SungWoo; Lee, Mi-Sun; Lieber, Charles M.

2013-01-01

Encoding electronic functionality into nanoscale elements during chemical synthesis has been extensively explored over the past decade as the key to developing integrated nanosystems1 with functions defined by synthesis2-6. Graphene7-12 has been recently explored as a two-dimensional nanoscale material, and has demonstrated simple device functions based on conventional top-down fabrication13-20. However, the synthetic approach to encoding electronic functionality and thus enabling an entire integrated graphene electronics in a chemical synthesis had not previously been demonstrated. Here we report an unconventional approach for the synthesis of monolithically-integrated electronic devices based on graphene and graphite. Spatial patterning of heterogeneous catalyst metals permits the selective growth of graphene and graphite, with controlled number of graphene layers. Graphene transistor arrays with graphitic electrodes and interconnects were formed from synthesis. These functional, all-carbon structures were transferrable onto a variety of substrates. The integrated transistor arrays were used to demonstrate real-time, multiplexed chemical sensing, and more significantly, multiple carbon layers of the graphene-graphite device components were vertically assembled to form a three-dimensional flexible structure which served as a top-gate transistor array. These results represent a substantial progress towards encoding electronic functionality via chemical synthesis and suggest future promise for one-step integration of graphene-graphite based electronics. PMID:22101813

Synthesis of monolithic graphene-graphite integrated electronics.

PubMed

Park, Jang-Ung; Nam, SungWoo; Lee, Mi-Sun; Lieber, Charles M

2011-11-20

Encoding electronic functionality into nanoscale elements during chemical synthesis has been extensively explored over the past decade as the key to developing integrated nanosystems with functions defined by synthesis. Graphene has been recently explored as a two-dimensional nanoscale material, and has demonstrated simple device functions based on conventional top-down fabrication. However, the synthetic approach to encoding electronic functionality and thus enabling an entire integrated graphene electronics in a chemical synthesis had not previously been demonstrated. Here we report an unconventional approach for the synthesis of monolithically integrated electronic devices based on graphene and graphite. Spatial patterning of heterogeneous metal catalysts permits the selective growth of graphene and graphite, with a controlled number of graphene layers. Graphene transistor arrays with graphitic electrodes and interconnects were formed from the synthesis. These functional, all-carbon structures were transferable onto a variety of substrates. The integrated transistor arrays were used to demonstrate real-time, multiplexed chemical sensing and more significantly, multiple carbon layers of the graphene-graphite device components were vertically assembled to form a three-dimensional flexible structure which served as a top-gate transistor array. These results represent substantial progress towards encoding electronic functionality through chemical synthesis and suggest the future promise of one-step integration of graphene-graphite based electronics.

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

Häusler, I., E-mail: ines.haeusler@bam.de; Dörfel, I., E-mail: Ilona.doerfel@bam.de; Peplinski, B., E-mail: Burkhard.peplinski@bam.de

A model system was used to simulate the properties of tribofilms which form during automotive braking. The model system was prepared by ball milling of a blend of 70 vol.% iron oxides, 15 vol.% molybdenum disulfide and 15 vol.% graphite. The resulting mixture was characterized by X-ray powder diffraction (XRD), X-ray photoelectron spectroscopy (XPS), and various transmission electron microscopic (TEM) methods, including energy dispersive X-ray spectroscopy (EDXS), high resolution investigations (HRTEM) with corresponding simulation of the HRTEM images, diffraction methods such as scanning nano-beam electron diffraction (SNBED) and selected area electron diffraction (SAED). It could be shown that the ballmore » milling caused a reduction of the grain size of the initial components to the nanometer range. Sometimes even amorphization or partial break-down of the crystal structure was observed for MoS{sub 2} and graphite. Moreover, chemical reactions lead to a formation of surface coverings of the nanoparticles by amorphous material, molybdenum oxides, and iron sulfates as derived from XPS. - Highlights: • Ball milling of iron oxides, MoS{sub 2}, and graphite to simulate a tribofilm • Increasing coefficient of friction after ball milling of the model blend • Drastically change of the diffraction pattern of the powder mixture • TEM & XPS showed the components of the milled mixture and the process during milling. • MoS{sub 2} and graphite suffered a loss in translation symmetry or became amorphous.« less

Structural Analysis of Pyrolytic Graphite Optics for the HiPEP Ion Thruster

NASA Technical Reports Server (NTRS)

Meckel, Nicole; Polaha, Jonathan; Juhlin, Nils

2006-01-01

The long lifetime requirements of interplanetary exploration missions is driving the need to develop long-life components for the electric propulsion thrusters that are being targeted for these missions. One of the primary life-limiting components of ion thrusters are the optics, which are continuously eroded during the operation of the thruster. Pyrolytic graphite optics are being considered for the High Power Electric Propulsion (HiPEP) ion thruster because of their very high resistance to erosion. This paper describes the structural analysis of the HiPEP pyrolytic graphite. A description of the development of the grid model, as well as the development of the effective properties and stress concentrations in the apertured area of the grids is included. An evaluation of the use of curved grids shows that the increased stiffness (compared to flat grids) prevents intergrid impact during launch, however, the residual stresses introduced by curving the grids pushes the resulting peak stresses beyond the critical stress. As a result, flat grids are recommended as the design solution. Thermally induced grid displacements during normal thruster operation are also presented.

High-temperature, high-pressure bonding of nested tubular metallic components

DOEpatents

Quinby, Thomas C.

1980-01-01

This invention is a tool for effecting high-temperature, high-compression bonding between the confronting faces of nested, tubular, metallic components. In a typical application, the tool is used to produce tubular target assemblies for irradiation in nuclear reactors or particle accelerators, the target assembly comprising a uranium foil and an aluminum-alloy substrate. The tool preferably is composed throughout of graphite. It comprises a tubular restraining member in which a mechanically expandable tubular core is mounted to form an annulus with the member. The components to be bonded are mounted in nested relation in the annulus. The expandable core is formed of individually movable, axially elongated segments whose outer faces cooperatively define a cylindrical pressing surface and whose inner faces cooperatively define two opposed, inwardly tapered, axial bores. Tapered rams extend respectively into the bores. The loaded tool is mounted in a conventional hot-press provided with evacuation means, heaters for maintaining its interior at bonding temperature, and hydraulic cylinders for maintaining a selected inwardly directed pressure on the tapered rams. With the hot-press evacuated and the loaded tool at the desired temperature, the cylinders are actuated to apply the selected pressure to the rams. The rams in turn expand the segmented core to maintain the nested components in compression against the restraining member. These conditions are maintained until the confronting faces of the nested components are joined in a continuous, uniform bond characterized by high thermal conductivity.

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)

Investigating the Co-Adsorption Behavior of Nucleic-Acid Base (Thymine and Cytosine) and Melamine at Liquid/Solid Interface

NASA Astrophysics Data System (ADS)

Zhao, Huiling; Li, Yinli; Chen, Dong; Liu, Bo

2016-12-01

The co-adsorption behavior of nucleic-acid base (thymine; cytosine) and melamine was investigated by scanning tunneling microscopy (STM) technique at liquid/solid (1-octanol/graphite) interface. STM characterization results indicate that phase separation happened after dropping the mixed solution of thymine-melamine onto highly oriented pyrolytic graphite (HOPG) surface, while the hetero-component cluster-like structure was observed when cytosine-melamine binary assembly system is used. From the viewpoints of non-covalent interactions calculated by using density functional theory (DFT) method, the formation mechanisms of these assembled structures were explored in detail. This work will supply a methodology to design the supramolecular assembled structures and the hetero-component materials composed by biological and chemical compound.

Design and Optimization of Composite Gyroscope Momentum Wheel Rings

NASA Technical Reports Server (NTRS)

Bednarcyk, Brett A.; Arnold, Steven M.

2007-01-01

Stress analysis and preliminary design/optimization procedures are presented for gyroscope momentum wheel rings composed of metallic, metal matrix composite, and polymer matrix composite materials. The design of these components involves simultaneously minimizing both true part volume and mass, while maximizing angular momentum. The stress analysis results are combined with an anisotropic failure criterion to formulate a new sizing procedure that provides considerable insight into the design of gyroscope momentum wheel ring components. Results compare the performance of two optimized metallic designs, an optimized SiC/Ti composite design, and an optimized graphite/epoxy composite design. The graphite/epoxy design appears to be far superior to the competitors considered unless a much greater premium is placed on volume efficiency compared to mass efficiency.

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.

Production of Metal-Free Composites Composed of Graphite Oxide and Oxidized Carbon Nitride Nanodots and Their Enhanced Photocatalytic Performances.

PubMed

Kim, Seung Yeon; Oh, Junghoon; Park, Sunghee; Shim, Yeonjun; Park, Sungjin

2016-04-04

A novel metal-free composite (GN) composed of two types of carbon-based nanomaterials, graphite oxide (GO) and 2D oxidized carbon nitride (OCN) nanodots was produced. Chemical and morphological characterizations reveal that GN contains a main component of GO with well-dispersed 2D OCN nanodots. GN shows enhanced photocatalytic performance for degrading an organic pollutant, Rhodamine B, under visible light. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Analytical techniques and instrumentation, a compilation

NASA Technical Reports Server (NTRS)

1974-01-01

Procedures for conducting materials tests and structural analyses of aerospace components are presented as a part of the NASA technology utilization program. Some of the subjects discussed are as follows: (1) failures in cryogenic tank insulation, (2) friction characteristics of graphite and graphite-metal combinations, (3) evaluation of polymeric products in thermal-vacuum environment, (4) erosion of metals by multiple impacts with water, (5) mass loading effects on vibrated ring and shell structures, (6) nonlinear damping in structures, and (7) method for estimating reliability of randomly excited structures.

The viscoelastic behavior of the principal compliance matrix of a unidirectional graphite/epoxy composite

NASA Technical Reports Server (NTRS)

Morris, D. H.; Yeow, Y. T.

1979-01-01

The time-temperature response of the principal compliances of a unidirectional graphite/epoxy composite was determined. It is shown that two components of the compliance matrix are time and temperature independent and that the compliance matrix is symmetric for the viscoelastic composite. The time-temperature superposition principle is used to determine shift factors which are independent of fiber orientation, for fiber angles that vary from 10 D to 90 D with respect to the load direction.

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

Technical Aspects Regarding the Management of Radioactive Waste from Decommissioning of Nuclear Facilities

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

Dragolici, F.; Turcanu, C. N.; Rotarescu, G.

2003-02-25

The proper application of the nuclear techniques and technologies in Romania started in 1957, once with the commissioning of the Research Reactor VVR-S from IFIN-HH-Magurele. During the last 45 years, appear thousands of nuclear application units with extremely diverse profiles (research, biology, medicine, education, agriculture, transport, all types of industry) which used different nuclear facilities containing radioactive sources and generating a great variety of radioactive waste during the decommissioning after the operation lifetime is accomplished. A new aspect appears by the planning of VVR-S Research Reactor decommissioning which will be a new source of radioactive waste generated by decontamination, disassemblingmore » and demolition activities. By construction and exploitation of the Radioactive Waste Treatment Plant (STDR)--Magurele and the National Repository for Low and Intermediate Radioactive Waste (DNDR)--Baita, Bihor county, in Romania was solved the management of radioactive wastes arising from operation and decommissioning of small nuclear facilities, being assured the protection of the people and environment. The present paper makes a review of the present technical status of the Romanian waste management facilities, especially raising on treatment capabilities of ''problem'' wastes such as Ra-266, Pu-238, Am-241 Co-60, Co-57, Sr-90, Cs-137 sealed sources from industrial, research and medical applications. Also, contain a preliminary estimation of quantities and types of wastes, which would result during the decommissioning project of the VVR-S Research Reactor from IFIN-HH giving attention to some special category of wastes like aluminum, graphite and equipment, components and structures that became radioactive through neutron activation. After analyzing the technical and scientific potential of STDR and DNDR to handle big amounts of wastes resulting from the decommissioning of VVR-S Research Reactor and small nuclear facilities, the necessity of up-gradation of these nuclear objectives before starting the decommissioning plan is revealed. A short presentation of the up-grading needs is also presented.« less

Research on graphite reinforced glass matrix composites

NASA Technical Reports Server (NTRS)

Prewo, K. M.; Thompson, E. R.

1980-01-01

High levels of mechanical performance in tension, flexure, fatigue, and creep loading situations of graphite fiber reinforced glass matrix composites are discussed. At test temperatures of up to 813 K it was found that the major limiting factor was the oxidative instability of the reinforcing graphite fibers. Particular points to note include the following: (1) a wide variety of graphite fibers were found to be comparable with the glass matrix composite fabrication process; (2) choice of fiber, to a large extent, controlled resultant composite performance; (3) composite fatigue performance was found to be excellent at both 300 K and 703 K; (4) composite creep and stress rupture at temperatures of up to 813 K was limited by the oxidative stability of the fiber; (5) exceptionally low values of composite thermal expansion coefficient were attributable to the dimensional stability of both matrix and fiber; and (6) component fabricability was demonstrated through the hot pressing of hot sections and brazing using glass and metal joining phases.

Bimetallic Metal-Organic Frameworks for Controlled Catalytic Graphitization of Nanoporous Carbons

PubMed Central

Tang, Jing; Salunkhe, Rahul R.; Zhang, Huabin; Malgras, Victor; Ahamad, Tansir; Alshehri, Saad M.; Kobayashi, Naoya; Tominaka, Satoshi; Ide, Yusuke; Kim, Jung Ho; Yamauchi, Yusuke

2016-01-01

Single metal-organic frameworks (MOFs), constructed from the coordination between one-fold metal ions and organic linkers, show limited functionalities when used as precursors for nanoporous carbon materials. Herein, we propose to merge the advantages of zinc and cobalt metals ions into one single MOF crystal (i.e., bimetallic MOFs). The organic linkers that coordinate with cobalt ions tend to yield graphitic carbons after carbonization, unlike those bridging with zinc ions, due to the controlled catalytic graphitization by the cobalt nanoparticles. In this work, we demonstrate a feasible method to achieve nanoporous carbon materials with tailored properties, including specific surface area, pore size distribution, degree of graphitization, and content of heteroatoms. The bimetallic-MOF-derived nanoporous carbon are systematically characterized, highlighting the importance of precisely controlling the properties of the carbon materials. This can be done by finely tuning the components in the bimetallic MOF precursors, and thus designing optimal carbon materials for specific applications. PMID:27471193

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.

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.

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

WE-AB-BRB-01: Development of a Probe-Format Graphite Calorimeter for Practical Clinical Dosimetry: Numerical Design Optimization, Prototyping, and Experimental Proof-Of-Concept

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

Renaud, J; Seuntjens, J; Sarfehnia, A

2015-06-15

Purpose: In this work, the feasibility of performing absolute dose to water measurements using a constant temperature graphite probe calorimeter (GPC) in a clinical environment is established. Methods: A numerical design optimization study was conducted by simulating the heat transfer in the GPC resulting from irradiation using a finite element method software package. The choice of device shape, dimensions, and materials was made to minimize the heat loss in the sensitive volume of the GPC. The resulting design, which incorporates a novel aerogel-based thermal insulator, and 15 temperature sensitive resistors capable of both Joule heating and measuring temperature, was constructedmore » in house. A software based process controller was developed to stabilize the temperatures of the GPC’s constituent graphite components to within a few 10’s of µK. This control system enables the GPC to operate in either the quasi-adiabatic or isothermal mode, two well-known, and independent calorimetry techniques. Absorbed dose to water measurements were made using these two methods under standard conditions in a 6 MV 1000 MU/min photon beam and subsequently compared against TG-51 derived values. Results: Compared to an expected dose to water of 76.9 cGy/100 MU, the average GPC-measured doses were 76.5 ± 0.5 and 76.9 ± 0.5 cGy/100 MU for the adiabatic and isothermal modes, respectively. The Monte Carlo calculated graphite to water dose conversion was 1.013, and the adiabatic heat loss correction was 1.003. With an overall uncertainty of about 1%, the most significant contributions were the specific heat capacity (type B, 0.8%) and the repeatability (type A, 0.6%). Conclusion: While the quasi-adiabatic mode of operation had been validated in previous work, this is the first time that the GPC has been successfully used isothermally. This proof-of-concept will serve as the basis for further study into the GPC’s application to small fields and MRI-linac dosimetry. This work has been supported in part by the CREATE Medical Physics Research Training Network of the Natural Sciences and Engineering Research Council (NSERC) grant 432290, NSERC grants RGPIN 298191 & 435608-13, Canadian Institutes of Health Research doctoral scholarship GSD-121793. This work has also been supported by Sun Nuclear Corporation.« less

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.

Multi-Scale Modeling of a Graphite-Epoxy-Nanotube System

NASA Technical Reports Server (NTRS)

Frankland, S. J. V.; Riddick, J. C.; Gates, T. S.

2005-01-01

A multi-scale method is utilized to determine some of the constitutive properties of a three component graphite-epoxy-nanotube system. This system is of interest because carbon nanotubes have been proposed as stiffening and toughening agents in the interlaminar regions of carbon fiber/epoxy laminates. The multi-scale method uses molecular dynamics simulation and equivalent-continuum modeling to compute three of the elastic constants of the graphite-epoxy-nanotube system: C11, C22, and C33. The 1-direction is along the nanotube axis, and the graphene sheets lie in the 1-2 plane. It was found that the C11 is only 4% larger than the C22. The nanotube therefore does have a small, but positive effect on the constitutive properties in the interlaminar region.

Diffusivity and solubility of hydrogen in the carbon fibre composite SEP N11

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

Alberici, S.; Perujo, A.; Camposilvan, J.

1995-10-01

In this paper we present the hydrogen diffusivity and solubility in the carbon fibre composite (CFC) SEP N11 with tri-directional fibres structure that is a possible candidate as armour material for plasma facing components (PFC). The technique used for these measurements is a gas evolution method and the measurements were carried out in the temperature range 900 - 1200 K with a loading hydrogen pressure of 100 kPa. The results obtained showed that the Sieverts` constant K{sub s} is of the same order of magnitude as those previously obtained for several graphites, while the diffusivity is about five to sixmore » orders of magnitude higher as compared to graphites. Furthermore, CFC presents an endothermic behaviour in contrast to graphites. 10 refs., 3 figs.« less

Encapsulation of α-Fe2O3 nanoparticles in graphitic carbon microspheres as high-performance anode materials for lithium-ion batteries

NASA Astrophysics Data System (ADS)

Zhang, Hongwei; Sun, Xiaoran; Huang, Xiaodan; Zhou, Liang

2015-02-01

A novel ``spray drying-carbonization-oxidation'' strategy has been developed for the fabrication of α-Fe2O3-graphitic carbon (α-Fe2O3@GC) composite microspheres, in which α-Fe2O3 nanoparticles with sizes of 30-50 nm are well-encapsulated by onion-like graphitic carbon shells with a thickness of 5-10 nm. In the constructed composite, the α-Fe2O3 nanoparticles act as the primary active material, providing a high capacity. Meanwhile, the graphitic carbon shells serve as the secondary active component, structural stabilizer, interfacial stabilizer, and electron-highway. As a result, the synthesized α-Fe2O3@GC nanocomposite exhibits a superior lithium-ion battery performance with a high reversible capacity (898 mA h g-1 at 400 mA g-1), outstanding rate capability, and excellent cycling stability. Our product, in terms of the facile and scalable preparation process and excellent electrochemical performance, demonstrates its great potential as a high-performance anode material for lithium-ion batteries.A novel ``spray drying-carbonization-oxidation'' strategy has been developed for the fabrication of α-Fe2O3-graphitic carbon (α-Fe2O3@GC) composite microspheres, in which α-Fe2O3 nanoparticles with sizes of 30-50 nm are well-encapsulated by onion-like graphitic carbon shells with a thickness of 5-10 nm. In the constructed composite, the α-Fe2O3 nanoparticles act as the primary active material, providing a high capacity. Meanwhile, the graphitic carbon shells serve as the secondary active component, structural stabilizer, interfacial stabilizer, and electron-highway. As a result, the synthesized α-Fe2O3@GC nanocomposite exhibits a superior lithium-ion battery performance with a high reversible capacity (898 mA h g-1 at 400 mA g-1), outstanding rate capability, and excellent cycling stability. Our product, in terms of the facile and scalable preparation process and excellent electrochemical performance, demonstrates its great potential as a high-performance anode material for lithium-ion batteries. Electronic supplementary information (ESI) available: XRD pattern, XPS spectrum, CV curves, TEM and SEM images, and table. See DOI: 10.1039/c4nr06771a

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

Utilization of the Philippine Research Reactor as a training facility for nuclear power plant operators

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

Palabrica, R.J.

1981-01-01

The Philippines has a 1-MW swimming-pool reactor facility operated by the Philippine Atomic Energy Commission (PAEC). The reactor is light-water moderated and cooled, graphite reflected, and fueled with 90% enriched uranium. Since it became critical in 1963 it has been utilized for research, radioisotope production, and training. It was used initially in the training of PAEC personnel and other research institutions and universities. During the last few years, however, it has played a key role in training personnel for the Philippine Nuclear Power Project (PNPP).

Develop, demonstrate, and verify large area composite structural bonding with polyimide adhesives. [adhesively bonding graphite-polyimide structures

NASA Technical Reports Server (NTRS)

Bhombal, B. D.; Wykes, D. H.; Hong, K. C.; Stenersen, A. A.

1982-01-01

The technology required to produce graphite-polyimide structural components with operational capability at 598 K (600 F) is considered. A series of polyimide adhesives was screened for mechanical and physical properties and processibility in fabricating large midplane bonded panels and honeycomb sandwich panels in an effort to fabricate a structural test component of the space shuttle aft body flap. From 41 formulations, LaRC-13, FM34B-18, and a modified LaRC-13 adhesive were selected for further evaluation. The LaRC-13 adhesive was rated as the best of the three adhesives in terms of availability, cost, processibility, properties, and ability to produce void fee large area (12" x 12") midplane bonds. Surface treatments and primers for the adhesives were evaluated and processes were developed for the fabrication of honeycomb sandwich panels of very good quality which was evidenced by rupture in the honeycomb core rather than in the facesheet bands on flatwise tensile strength testing. The fabrication of the adhesively bonded honeycomb sandwich cover panels, ribs, and leading edge covers of Celion graphite/LARC-160 polyimide laminates is described.

PhybalSIT — Fatigue Assessment and Life Time Calculation of the Ductile Cast Iron EN-GJS-600 at Ambient and Elevated Temperatures

NASA Astrophysics Data System (ADS)

Jost, Benjamin; Klein, Marcus; Eifler, Dietmar

This paper focuses on the ductile cast iron EN-GJS-600 which is often used for components of combustion engines. Under service conditions, those components are mechanically loaded at different temperatures. Therefore, this investigation targets at the fatigue behavior of EN-GJS-600 at ambient and elevated temperatures. Light and scanning electron microscopic investigations were done to characterize the sphericity of the graphite as well as the ferrite, pearlite and graphite fraction. At elevated temperatures, the consideration of dynamic strain ageing effects is of major importance. In total strain increase, temperature increase and constant total strain amplitude tests, the plastic strain amplitude, the stress amplitude, the change in temperature and the change in electrical resistance were measured. The measured values depend on plastic deformation processes in the bulk of the specimens and at the interfaces between matrix and graphite. The fatigue behavior of EN-GJS-600 is dominated by cyclic hardening processes. The physically based fatigue life calculation "PHYBALSIT" (SIT = strain increase test) was developed for total strain controlled fatigue tests. Only one temperature increase test is necessary to determine the temperature interval of pronounced dynamic strain ageing effects.

Comparison of Fluka-2006 Monte Carlo Simulation and Flight Data for the ATIC Detector

NASA Technical Reports Server (NTRS)

Gunasingha, R.M.; Fazely, A.R.; Adams, J.H.; Ahn, H.S.; Bashindzhagyan, G.L.; Chang, J.; Christl, M.; Ganel, O.; Guzik, T.G.; Isbert, J.;

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

Reproducibility of structural strength and stiffness for graphite-epoxy aircraft spoilers

NASA Technical Reports Server (NTRS)

Howell, W. E.; Reese, C. D.

1978-01-01

Structural strength reproducibility of graphite epoxy composite spoilers for the Boeing 737 aircraft was evaluated by statically loading fifteen spoilers to failure at conditions simulating aerodynamic loads. Spoiler strength and stiffness data were statistically modeled using a two parameter Weibull distribution function. Shape parameter values calculated for the composite spoiler strength and stiffness were within the range of corresponding shape parameter values calculated for material property data of composite laminates. This agreement showed that reproducibility of full scale component structural properties was within the reproducibility range of data from material property tests.

Development and demonstration of manufacturing processes for fabricating graphite/Larc-160 polyimide structural elements, part 4, paragraph B

NASA Technical Reports Server (NTRS)

1981-01-01

Progress in the development of processes for production of Celion/LARC-160 graphite-polyimide materials, quality control, and the fabrication of Space Shuttle composite structure components is reported. Liquid chromatographic analyses of three repeatibility batches were performed and are compared to previous Hexcel standard production and to variables study LARC-160 intermediate resins. Development of processes for chopped fiber molding are described and flexural strength, elastic modulus, and other physical and mechanical properties of the molding are presented.

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)

A review on management of spent lithium ion batteries and strategy for resource recycling of all components from them.

PubMed

Zhang, Wenxuan; Xu, Chengjian; He, Wenzhi; Li, Guangming; Huang, Juwen

2018-02-01

The wide use of lithium ion batteries (LIBs) has brought great numbers of discarded LIBs, which has become a common problem facing the world. In view of the deleterious effects of spent LIBs on the environment and the contained valuable materials that can be reused, much effort in many countries has been made to manage waste LIBs, and many technologies have been developed to recycle waste LIBs and eliminate environmental risks. As a review article, this paper introduces the situation of waste LIB management in some developed countries and in China, and reviews separation technologies of electrode components and refining technologies of LiCoO 2 and graphite. Based on the analysis of these recycling technologies and the structure and components characteristics of the whole LIB, this paper presents a recycling strategy for all components from obsolete LIBs, including discharge, dismantling, and classification, separation of electrode components and refining of LiCoO 2 /graphite. This paper is intended to provide a valuable reference for the management, scientific research, and industrial implementation on spent LIBs recycling, to recycle all valuable components and reduce the environmental pollution, so as to realize the win-win situation of economic and environmental benefits.

Dual-Layer Oxidation-Protective Plasma-Sprayed SiC-ZrB2/Al2O3-Carbon Nanotube Coating on Graphite

NASA Astrophysics Data System (ADS)

Ariharan, S.; Sengupta, Pradyut; Nisar, Ambreen; Agnihotri, Ankur; Balaji, N.; Aruna, S. T.; Balani, Kantesh

2017-02-01

Graphite is used in high-temperature gas-cooled reactors because of its outstanding irradiation performance and corrosion resistance. To restrict its high-temperature (>873 K) oxidation, atmospheric-plasma-sprayed SiC-ZrB2-Al2O3-carbon nanotube (CNT) dual-layer coating was deposited on graphite substrate in this work. The effect of each layer was isolated by processing each component of the coating via spark plasma sintering followed by isothermal kinetic studies. Based on isothermal analysis and the presence of high residual thermal stress in the oxide scale, degradation appeared to be more severe in composites reinforced with CNTs. To avoid the complexity of analysis of composites, the high-temperature activation energy for oxidation was calculated for the single-phase materials only, yielding values of 11.8, 20.5, 43.5, and 4.5 kJ/mol for graphite, SiC, ZrB2, and CNT, respectively, with increased thermal stability for ZrB2 and SiC. These results were then used to evaluate the oxidation rate for the composites analytically. This study has broad implications for wider use of dual-layer (SiC-ZrB2/Al2O3) coatings for protecting graphite crucibles even at temperatures above 1073 K.

High-resolution transmission electron microscopy studies of graphite materials prepared by high-temperature treatment of unburned carbon concentrates from combustion fly ashes

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

Miguel Cabielles; Jean-Nol Rouzaud; Ana B. Garcia

2009-01-15

High-resolution transmission electron microscopy (HRTEM) has been used in this work to study the microstructural (structure and microtexture) changes occurring during the high-temperature treatment of the unburned carbon concentrates from coal combustion fly ashes. Emphasis was placed on two aspects: (i) the development of graphitic carbon structures and (ii) the disordered carbon forms remaining in the graphitized samples. In addition, by coupling HRTEM with energy-dispersive spectroscopy, the transformations with the temperature of the inorganic matter (mainly iron- and silicon-based phases) of the unburned carbon concentrates were evidenced. The HRTEM results were compared to the averaged structural order of the materialsmore » as evaluated by X-ray diffraction (XRD) and Raman spectroscopy. As indicated by XRD and Raman parameters, more-ordered materials were obtained from the unburned carbon concentrates with higher mineral/inorganic matter, thus inferring the catalytic effect of some of their components. However, the average character of the information provided by these instrumental techniques seems to be inconclusive in discriminating between carbon structures with different degrees of order (stricto sensu graphite, graphitic, turbostratic, etc.) in a given graphitized unburned carbon. Unlike XRD and Raman, HRTEM is a useful tool for imaging directly the profile of the polyaromatic layers (graphene planes), thus allowing the sample heterogeneity to be looked at, specifically the presence of disordered carbon phases. 49 refs., 9 figs., 3 tabs.« less

Graphite Fiber Textile Preform/Copper Matrix Composites

NASA Technical Reports Server (NTRS)

Filatovs, G. J.; Lee, Bruce; Bass, Lowell

1996-01-01

Graphite fiber reinforced/copper matrix composites are candidate materials for critical heat transmitting and rejection components because of their high thermal conduction. The use of textile (braid) preforms allows near-net shapes which confers additional advantages, both for enhanced thermal conduction and increased robustness of the preform against infiltration and handling damage. Issues addressed in the past year center on the determination of the braid structure following infiltration, and the braidability vs. the conductivity of the fibers. Highly conductive fibers eventuate from increased graphitization, which increases the elastic modulus, but lowers the braidability; a balance between these factors must be achieved. Good quality braided preform bars have been fabricated and infiltrated, and their thermal expansion characterized; their analytic modeling is underway. The braided preform of an integral finned tube has been fabricated and is being prepared for infiltration.

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.

Swift heavy ion-induced radiation damage in isotropic graphite studied by micro-indentation and in-situ electrical resistivity

NASA Astrophysics Data System (ADS)

Hubert, Christian; Voss, Kay Obbe; Bender, Markus; Kupka, Katharina; Romanenko, Anton; Severin, Daniel; Trautmann, Christina; Tomut, Marilena

2015-12-01

Due to its excellent thermo-physical properties and radiation hardness, isotropic graphite is presently the most promising material candidate for new high-power ion accelerators which will provide highest beam intensities and energies. Under these extreme conditions, specific accelerator components including production targets and beam protection modules are facing the risk of degradation due to radiation damage. Ion-beam induced damage effects were tested by irradiating polycrystalline, isotropic graphite samples at the UNILAC (GSI, Darmstadt) with 4.8 MeV per nucleon 132Xe, 150Sm, 197Au, and 238U ions applying fluences between 1 × 1011 and 1 × 1014 ions/cm2. The overall damage accumulation and its dependence on energy loss of the ions were studied by in situ 4-point resistivity measurements. With increasing fluence, the electric resistivity increases due to disordering of the graphitic structure. Irradiated samples were also analyzed off-line by means of micro-indentation in order to characterize mesoscale effects such as beam-induced hardening and stress fields within the specimen. With increasing fluence and energy loss, hardening becomes more pronounced.

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.

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

Paul Demkowicz; Lance Cole; Scott Ploger

The AGR-1 irradiation experiment ended on November 6, 2009, after 620 effective full power days in the Advanced Test Reactor, achieving a peak burnup of 19.6% FIMA. The test train was shipped to the Materials and Fuels Complex in March 2010 for post-irradiation examination. The first PIE activities included non-destructive examination of the test train, followed by disassembly of the test train and individual capsules and detailed inspection of the capsule contents, including the fuel compacts and the graphite fuel holders. Dimensional measurements of the compacts, graphite holders, and steel capsules shells were performed using a custom vision measurement systemmore » (for outer diameters and lengths) and conventional bore gauges (for inner diameters). Gamma spectrometry of the intact test train gave a preliminary look at the condition of the interior components. No evidence of damage to compacts or graphite components was evident from the isotopic and gross gamma scans. Neutron radiography of the intact Capsule 2 showed a high degree of detail of interior components and confirmed the observation that there was no major damage to the capsule. Disassembly of the capsules was initiated using procedures qualified during out-of-cell mockup testing. Difficulties were encountered during capsule disassembly due to irradiation-induced changes in some of the capsule components’ properties, including embrittled niobium and molybdenum parts that were susceptible to fracture and swelling of the graphite fuel holders that affected their removal from the capsule shells. This required various improvised modifications to the disassembly procedure to avoid damage to the fuel compacts. Ultimately the capsule disassembly was successful and only one compact from Capsule 4 (out of 72 total in the test train) sustained damage during the disassembly process, along with the associated graphite holder. The compacts were generally in very good condition upon removal. Only relatively minor damage or markings were visible using high resolution photographic inspection. Compact dimensional measurements indicated diametrical shrinkage of 0.9 to 1. 4%, and length shrinkage of 0.2 to 1.1%. The shrinkage was somewhat dependent on compact location within each capsule and within the test train. Compacts exhibited a maximum diametrical shrinkage at a fast neutron fluence of approximately 3×1021 n/cm2. A multivariate statistical analysis indicates that fast neutron fluence as well as compact position in the test train influence compact shrinkage.« less

Supernova graphite in the NanoSIMS: Carbon, oxygen and titanium isotopic compositions of a spherule and its TiC sub-components

NASA Astrophysics Data System (ADS)

Stadermann, F. J.; Croat, T. K.; Bernatowicz, T. J.; Amari, S.; Messenger, S.; Walker, R. M.; Zinner, E.

2005-01-01

Presolar graphite spherules from the Murchison low-density separate KE3 contain a large number of internal TiC crystals that range in size from 15 to 500 nm. We have studied one such graphite grain in great detail by successive analyses with SEM, ims3f SIMS, TEM and NanoSIMS. Isotopic measurements of the 'bulk' particle in the ims3f indicate a supernova origin for this graphite spherule. The NanoSIMS measurements of C, N, O and Ti isotopes were performed directly on TEM ultramicrotome sections of the spherule, allowing correlated studies of the isotopic and mineralogical properties of the graphite grain and its internal crystals. We found isotopic gradients in 12C/ 13C and 16O/ 18O from the core of the graphite spherule to its perimeter, with the most anomalous compositions being present in the center. These gradients may be the result of isotopic exchange with isotopically normal material, either in the laboratory or during the particle's history. No similar isotopic gradients were found in the 16O/ 17O and 14N/ 15N ratios, which are normal within analytical uncertainty throughout the graphite spherule. Due to an unusually high O signal, internal TiC crystals were easily located during NanoSIMS imaging measurements. It was thus possible to determine isotopic compositions of several internal TiC grains independent of the surrounding graphite matrix. These TiC crystals are significantly more anomalous in their O isotopes than the graphite, with 16O/ 18O ratios ranging from 14 to 250 (compared to a terrestrial value of 499). Even the most centrally located TiC grains show significant variations in their O isotopic compositions from crystal to crystal. Measurement of the Ti isotopes in three TiC grains found no variations among them and no large differences between the compositions of the different crystals and the 'bulk' graphite spherule. However, the same three TiC crystals vary by a factor of 3 in their 16O/ 18O ratios. It is not clear in what form the O is associated with the TiC grains and whether it is cogenetic or the result of surface reactions on the TiC grains before they accreted onto the growing graphite spherule. The presence of 44Ca from short-lived 44Ti (t 1/2 = 60y) in one of the TiC subgrains confirms the identification of this graphite spherule as a supernova condensate.

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

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

Sodium-22 In Supernovae

NASA Astrophysics Data System (ADS)

Amari, Sachiko

2008-05-01

There are several isotopically distinct noble gas components in meteorites. Of them, Ne-E(L), heavily enriched in 22Ne, is carried by graphite with a range of density (1.6 - 2.2 g/cm3). Bulk (=aggregates) noble gas analysis of graphite separates from the Murchison meteorite indicate that a dominant source of 22Ne is 22Na (T1/2 = 2.6 a) with varying proportions of 22Ne via 14N(α,γ)18F(e+ν)18O(α,γ)22Ne with density. Low-density graphite grains, from their isotopic signatures, are believed to have formed in supernovae. Examinations of both bulk and single-grain analyses of low-density graphite grains (Amari et al., 1995; Nichols et al., 1994) indicate that all 22Ne in low-density graphite grains is from the decay of 22Na that was produced in the O/Ne zone in supernovae. One may argue why implanted 20,22Ne was not observed in the grains, considering the fact that the mass fraction of 20Ne is 5 orders of magnitude larger than that of 22Na. Croat et al. (2003) observed TiC subgrains inside low-density graphite grains have amorphous rims with the thickness of 3 to 15 nm, indicating atom bombardment from the surrounding gas. Assuming the gas is He, they estimated the velocity is 50 km/s or less. If the relative velocities between the Ne and the graphite grains are in that range, the penetration depth into the graphite grains is 2nm. Such shallow surface layers would be sputtered once the grains hit the reverse shock and keep traveling into the hot H-rich region (Nozawa et al, 2007). It remains to be seen whether or not 22Na in higher-density graphite is from supernovae or novae, or both. Amari, S. et al. 1995, Geochim. Cosmochim. Acta, 59, 1411 Croat, T.K. et al. 2003, Geochim. Cosmochim. Acta, 67, 4705 Nichols R.H. et al. 1994, Meteoritics, 29, 510 Nozawa, T. et al. 2007 ApJ, 666, 955

Tracing Life in the Earliest Terrestrial Rock Record

NASA Astrophysics Data System (ADS)

Lepland, A.; van Zuilen, M.; Arrhenius, G.

2001-12-01

The principal method for studying the earliest traces of life in the metamorphosed, oldest (> 3.5 Ga) terrestrial rocks involves determination of isotopic composition of carbon, mainly prevailing as graphite. It is generally believed that this measure can distinguish biogenic graphite from abiogenic varieties. However, the interpretation of life from carbon isotope ratios has to be assessed within the context of specific geologic circumstances requiring (i) reliable protolith interpretation (ii) control of secondary, metasomatic processes, and (iii) understanding of different graphite producing mechanisms and related carbon isotopic systematics. We have carried out a systematic study of abundance, isotopic composition and petrographic associations of graphite in rocks from the ca. 3.8 Ga Isua Supracrustal Belt (ISB) in southern West Greenland. Our study indicates that most of the graphite in ISB occurs in carbonate-rich metasomatic rocks (metacarbonates) while sedimentary units, including banded iron formations (BIFs) and metacherts, have exceedingly low graphite concentrations. Regardless of isotopic composition of graphite in metacarbonate rocks, their secondary origin disqualifies them from providing evidence for traces of life stemming from 3.8 Ga. Recognition of the secondary origin of Isua metacarbonates thus calls for reevaluation of biologic interpretations by Schidlowski et al. (1979) and Mojzsis et al. (1996) that suggested the occurrence of 3.8 Ga biogenic graphite in these rocks. The origin of minute quantities of reduced carbon, released from sedimentary BIFs and metacherts at combustion steps > 700 C remains to be clarified. Its isotopic composition (d13C from -18 to -25%) may hint at a biogenic origin. However, such isotopically light carbon was also found in Proterozoic mafic dykes cross-cutting the metasedimentary units in the ISB. The occurrence of isotopically light, reduced carbon in biologically irrelevant dykes may indicate secondary graphite crystallization from CO2 or CH4- containing fluids that in turn may derive from bioorganic sources. If this were the case, trace amounts of isotopically light secondary graphite can also be expected in metasediments, complicating the usage of light graphite as primary biomarker. The possibility of recent organic contamination, particularly important in low graphite samples, needs also to be considered; it appears as a ubiquitous component released at combustion in the 400 to 500 deg range. - A potential use of the apatite-graphite association as a biomarker has been proposed in the study by Mojzsis et al. (1996). Close inspection of several hundred apatite crystals from Isua BIFs and metacherts did, however, not show an association between these two minerals, moreover graphite is practically absent in these metasediments. In contrast, apatite crystals in the non-sedimentary metacarbonate rocks were found commonly to have invaginations, coatings and inclusions of abundant graphite. Considering that such graphite inclusions in apatite are restricted to the secondary metasomatic carbonate rocks in the ISB this association can not be considered as a primary biomarker in the Isua Supracrustal Belt References: Mojzsis,S.J, .Arrhenius,G., McKeegan, K.D.,.Harrison, T.M.,.Nutman, A.P & C.R.L.Friend.,1996. Nature 384: 55 Schidlowski, M., Appel, P.W.U., Eichmann, R. & Junge, C.E., 1979. Geochim. Cosmochim. Acta 43: 189-190.

Graphite fiber textile preform/copper matrix composites

NASA Technical Reports Server (NTRS)

Filatovs, G. J.

1993-01-01

This project has the objective of exploring the use of graphite fiber textile preform/copper matrix composites in spacecraft heat transmitting and radiating components. The preforms are to be fabricated by braiding of tows and when infiltrated with copper will result in a 3-D reinforced, near net shape composite with improved specific properties such as lower density and higher stiffness. It is anticipated that the use of textile technology will result in a more robust preform and consequently better final composite; it is hard to anticipate what performance tradeoffs will result, and these will be explored through testing and characterization.

Percolation transition in carbon composite on the basis of fullerenes and exfoliated graphite

NASA Astrophysics Data System (ADS)

Berezkin, V. I.; Popov, V. V.

2018-01-01

The electrical conductivity of a carbon composite on the basis of C60 fullerenes and exfoliated graphite is investigated in the range of relative contents of components from 0 to 100%. The samples are obtained by the thermal treatment of the initial dispersed mixtures in vacuum in the diffusion-adsorption process and their further cold pressing. The resistivity of the samples gradually increases with an increase in the fraction of fullerenes, and a sharp transition from the conductive state to the dielectric one is observed after achieving certain concentrations of C60. The interpretation of the results within the percolation theory makes it possible to evaluate the percolation threshold (expressed as a relative content of graphite) as equal to 4.45 wt % and the critical conductivity index as equal to 1.85 (which is typical for three-dimensional twocomponent disordered media including those having pores).

Rice husk-originating silicon-graphite composites for advanced lithium ion battery anodes.

PubMed

Kim, Hye Jin; Choi, Jin Hyeok; Choi, Jang Wook

2017-01-01

Rice husk is produced in a massive amount worldwide as a byproduct of rice cultivation. Rice husk contains approximately 20 wt% of mesoporous SiO 2 . We produce mesoporous silicon (Si) by reducing the rice husk-originating SiO 2 using a magnesio-milling process. Taking advantage of meso-porosity and large available quantity, we apply rice husk-originating Si to lithium ion battery anodes in a composite form with commercial graphite. By varying the mass ratio between these two components, trade-off relation between specific capacity and cycle life was observed. A controllable pre-lithiation scheme was adopted to increase the initial Coulombic efficiency and energy density. The series of electrochemical results suggest that rice husk-originating Si-graphite composites are promising candidates for high capacity lithium ion battery anodes, with the prominent advantages in battery performance and scalability.

A model of the thermal-spike mechanism in graphite/epoxy laminates

NASA Technical Reports Server (NTRS)

Adamson, M. J.

1982-01-01

The influence of a thermal spike on a moisture-saturated graphite/epoxy composite was studied in detail. A single thermal spike from 25 C to 132 C was found to produce damage as evidenced by a significant increase in the level of moisture saturation in the composite. Approximately half of this increase remained after a vacuum anneal at 150 C for 7 days, suggesting the presence of an irreversible damage component. Subsequent thermal spikes created less and less additional moisture absorption, with the cumulative effect being a maximum or limiting moisture capacity of the composite. These observations are explained in terms of a model previously developed to explain the reverse thermal effect of moisture absorption in epoxy and epoxy matrix composites. This model, based on the inverse temperature dependence of free volume, contributes an improved understanding of thermal-spike effects in graphite/epoxy composites.

Impact response of graphite-epoxy flat laminates using projectiles that simulate aircraft engine encounters

NASA Technical Reports Server (NTRS)

Preston, J. L., Jr.; Cook, T. S.

1975-01-01

An investigation of the response of a graphite-epoxy material to foreign object impact was made by impacting spherical projectiles of gelatin, ice, and steel normally on flat panels. The observed damage was classified as transverse (stress wave delamination and cracking), penetrative, or structural (gross failure): the minimum, or threshold, velocity to cause each class of damage was established as a function of projectile characteristics. Steel projectiles had the lowest transverse damage threshold, followed by gelatin and ice. Making use of the threshold velocities and assuming that the normal component of velocity produces the damage in nonnormal impacts, a set of impact angles and velocities was established for each projectile material which would result in damage to composite fan blades. Analysis of the operating parameters of a typical turbine fan blade shows that small steel projectiles are most likely to cause delamination and penetration damage to unprotected graphite-epoxy composite fan blades.

Direct manufacturing of ultrathin graphite on three-dimensional nanoscale features

PubMed Central

Pacios, Mercè; Hosseini, Peiman; Fan, Ye; He, Zhengyu; Krause, Oliver; Hutchison, John; Warner, Jamie H.; Bhaskaran, Harish

2016-01-01

There have been many successful attempts to grow high-quality large-area graphene on flat substrates. Doing so at the nanoscale has thus far been plagued by significant scalability problems, particularly because of the need for delicate transfer processes onto predefined features, which are necessarily low-yield processes and which can introduce undesirable residues. Herein we describe a highly scalable, clean and effective, in-situ method that uses thin film deposition techniques to directly grow on a continuous basis ultrathin graphite (uG) on uneven nanoscale surfaces. We then demonstrate that this is possible on a model system of atomic force probe tips of various radii. Further, we characterize the growth characteristics of this technique as well as the film’s superior conduction and lower adhesion at these scales. This sets the stage for such a process to allow the use of highly functional graphite in high-aspect-ratio nanoscale components. PMID:26939862

Ureilite smelting

NASA Technical Reports Server (NTRS)

Walker, David; Grove, Tim

1993-01-01

Ureilites containing homogeneous Fo76 olivine cores in intimate co-existence with graphite must have recrystallized at pressures of at least approximately 100 bars to suppress smelting of the fayalite component of the olivine to Fe metal. Smelting of olivine and pyroxene-saturated magmatic liquids produces orthopyroxene-without-olivine crystalline derivatives unlike those in ureilites. Thus the Mg# compositional variation within the ureilite suite, which is commonly attributed to partial smelting, cannot plausibly be produced by assemblages rich in liquid. In situ smelting of graphitic olivine + pigeonite crystal mushes can produce the correct crystal assemblage, but fails to provide a plausible account for the removal of metal from ureilites or for the correlation of Mg# with Delta O-17. Even if Mg# and Delta O-17 variations are established in the nebula, ureilite recrystallization with graphite must have occurred at pressures greater than the minima we have experimentally established, corresponding to parent objects not less than approximately 100 km in radius.

Lithium loss in the solid electrolyte interphase: Lithium quantification of aged lithium ion battery graphite electrodes by means of laser ablation inductively coupled plasma mass spectrometry and inductively coupled plasma optical emission spectroscopy

NASA Astrophysics Data System (ADS)

Schwieters, Timo; Evertz, Marco; Mense, Maximilian; Winter, Martin; Nowak, Sascha

2017-07-01

In this work we present a new method using LA-ICP-MS to quantitatively determine the lithium content in aged graphite electrodes of a lithium ion battery (LIB) by performing total depth profiling. Matrix matched solid external standards are prepared using a solid doping approach to avoid elemental fractionation effects during the measurement. The results are compared and matched to the established ICP-OES technique for bulk quantification after performing a microwave assisted acid digestion. The method is applied to aged graphite electrodes in order to determine the lithium immobilization (= "Li loss") in the solid electrolyte interphase after the first cycle of formation. For this, different samples including a reference sample are created to obtain varying thicknesses of the SEI covering the electrode particles. By applying defined charging voltages, an initial lithiation process is performed to obtain specific graphite intercalation compounds (GICs, with target stoichiometries of LiC30, LiC18, LiC12 and LiC6). Afterwards, the graphite electrode is completely discharged to obtain samples without mobile, thus active lithium in its lattice. Taking the amount of lithium into account which originates from the residues of the LiPF6 (dissolved in the carbon components containing electrolyte), it is possible to subtract the amount of lithium in the SEI.

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.].

COMPOSITION AND METHOD FOR COATING A CERAMIC BODY

DOEpatents

Blanchard, M.K.

1958-11-01

A method is presented for protecting a beryllium carbide-graphite body. The method consists in providing a ceramic coating which must contain at least one basic oxide component, such as CaO, at least one amphoteric oxide component, such as Al/sub 2/O/sub 3/, and at least one acidic oxide component, such as SiO/ sub 2/. Various specific formulations for this ceramic coating are given and the coating is applied by conventional ceramic techniques.

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.

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.

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.

BRAZING ALLOYS

DOEpatents

Donnelly, R.G.; Gilliland, R.G.; Slaughter, G.M.

1962-02-20

A brazing alloy is described which, in the molten state, is characterized by excellent wettability and flowability and is capable of forming a corrosion-resistant brazed joint. At least one component of said joint is graphite and the other component is a corrosion-resistant refractory metal. The brazing alloy consists essentially of 40 to 90 wt % of gold, 5 to 35 wt% of nickel, and 1 to 45 wt% of tantalum. (AEC)

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.

TUNGSTEN BASE ALLOYS

DOEpatents

Schell, D.H.; Sheinberg, H.

1959-12-15

A high-density quaternary tungsten-base alloy having high mechanical strength and good machinability composed of about 2 wt.% Ni, 3 wt.% Cu, 5 wt.% Pb, and 90wt.% W is described. This alloy can be formed by the powder metallurgy technique of hot pressing in a graphite die without causing a reaction between charge and the die and without formation of a carbide case on the final compact, thereby enabling re-use of the graphite die. The alloy is formable at hot- pressing temperatures of from about 1200 to about 1350 deg C. In addition, there is little component shrinkage, thereby eliminating the necessity of subsequent extensive surface machining.

IMPREGNATION OF GRAPHITE SPECIMENS USING THE FURFURYL ALCOHOL IMPREGNATION PROCESS. Project DRAGON

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

Martina, R.A.; Vohler, O.J.

1964-11-15

This report gives the results of the furfuryl alcohol impregnation treatment of graphite fuel rods and end plugs done by Sigri Kohlefabrikate GmbH, Meitingen, from May 1962 till April 1964. By impregnating the components two and three times, resp. the permeability was decreased, starting from a coefficient of about 5 x 10{sup -2}cm{sup 2}sec{sup -1}, to a value of 10{sup -6} - 10{sup -7}cm{sup 2}sec{sup -1}. The rejects as occured with the first batches were finally diminished to a negligible amount by studying the single steps of treatment. (auth)

Carbothermal transformation of a graphitic carbon nanofiber/silica aerogel composite to a SiC/silica nanocomposite.

PubMed

Lu, Weijie; Steigerwalt, Eve S; Moore, Joshua T; Sullivan, Lisa M; Collins, W Eugene; Lukehart, C M

2004-09-01

Carbon nanofiber/silica aerogel composites are prepared by sol-gel processing of surface-enhanced herringbone graphitic carbon nanofibers (GCNF) and Si(OMe)4, followed by supercritical CO2 drying. Heating the resulting GCNF/silica aerogel composites to 1650 degrees C under a partial pressure of Ar gas initiates carbothermal reaction between the silica aerogel matrix and the carbon nanofiber component to form SiC/silica nanocomposites. The SiC phase is present as nearly spherical nanoparticles, having an average diameter of ca. 8 nm. Formation of SiC is confirmed by powder XRD and by Raman spectroscopy.

Uncovering the local inelastic interactions during manufacture of ductile cast iron: How the substructure of the graphite particles can induce residual stress concentrations in the matrix

NASA Astrophysics Data System (ADS)

Andriollo, Tito; Hellström, Kristina; Sonne, Mads Rostgaard; Thorborg, Jesper; Tiedje, Niels; Hattel, Jesper

2018-02-01

Recent X-ray diffraction (XRD) measurements have revealed that plastic deformation and a residual elastic strain field can be present around the graphite particles in ductile cast iron after manufacturing, probably due to some local mismatch in thermal contraction. However, as only one component of the elastic strain tensor could be obtained from the XRD data, the shape and magnitude of the associated residual stress field have remained unknown. To compensate for this and to provide theoretical insight into this unexplored topic, a combined experimental-numerical approach is presented in this paper. First, a material equivalent to the ductile cast iron matrix is manufactured and subjected to dilatometric and high-temperature tensile tests. Subsequently, a two-scale hierarchical top-down model is devised, calibrated on the basis of the collected data and used to simulate the interaction between the graphite particles and the matrix during manufacturing of the industrial part considered in the XRD study. The model indicates that, besides the viscoplastic deformation of the matrix, the effect of the inelastic deformation of the graphite has to be considered to explain the magnitude of the XRD strain. Moreover, the model shows that the large elastic strain perturbations recorded with XRD close to the graphite-matrix interface are not artifacts due to e.g. sharp gradients in chemical composition, but correspond to residual stress concentrations induced by the conical sectors forming the internal structure of the graphite particles. In contrast to common belief, these results thus suggest that ductile cast iron parts cannot be considered, in general, as stress-free at the microstructural scale.

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.

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.

The Place for Thermoplastic Composites in Structural Components

DTIC Science & Technology

1987-12-01

The molten tube is then squashed flat and consolidated into ribbon form by continuous opposed-belt laminating. Existing graphite-epoxy pultrusion...the solid form it must have a molecular weight that exceeds the critical entanglement value. Thus thermoplastic materials of commercial worth almost

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

Development of CVD-W coatings on CuCrZr and graphite substrates with a PVD intermediate layer

NASA Astrophysics Data System (ADS)

Song, Jiupeng; Lian, Youyun; Lv, Yanwei; Liu, Junyong; Yu, Yang; Liu, Xiang; Yan, Binyou; Chen, Zhigang; Zhuang, Zhigang; Zhao, Ximeng; Qi, Yang

2014-12-01

In order to apply tungsten (W) coatings by chemical vapor deposition (CVD) for repairing or updating the plasma facing components (PFCs) of the first wall and divertor in existing or future tokomaks, where CuCrZr or graphite is the substrate material, an intermediate layer by physical vapor deposition (PVD) has been used to accommodate the interface stress due to the mismatch of thermal expansion or act as a diffusion barrier between the CVD-W coating and the substrate. The prepared CuCrZr/PVD-Cu/CVD-W sample with active cooling has passed thermal fatigue tests by electron beam with an absorbed power of 2.2 MW/m2, 50 s on/50 s off, for 100 cycles. Another graphite/PVD-Si/CVD-W sample without active cooling underwent thermal fatigue testing with an absorbed power density of 4.62 MW/m2, 5 s on/25 s off, for 200 cycles, and no catastrophic failure was found.

Systems integration and demonstration of advanced reusable structure for ALS

NASA Technical Reports Server (NTRS)

Gibbins, Martin N.

1991-01-01

The objective was to investigate the potential of advanced material to achieve life cycle cost (LCC) benefits for reusable structure on the advanced launch system. Three structural elements were investigated - all components of an Advanced Launch System reusable propulsion/avionics module. Leading aeroshell configurations included sandwich structure using titanium, graphite/polyimide (Gr/PI), or high-temperature aluminum (HTA) face sheets. Thrust structure truss concepts used titanium, graphite/epoxy, or silicon carbide/aluminum struts. Leading aft bulkhead concepts employed graphite epoxy and aluminum. The technical effort focused on the aeroshell because the greatest benefits were expected there. Thermal analyses show the structural temperature profiles during operation. Finite element analyses show stresses during splash-down. Weight statements and manufacturing cost estimates were prepared for calculation of LCC for each design. The Gr/PI aeroshell showed the lowest potential LCC, but the HTA aeroshell was judged to be lower risk. A technology development plan was prepared to validate the applicable structural technology.

Enhancement of gaps in thin graphitic films for heterostructure formation

NASA Astrophysics Data System (ADS)

Hague, J. P.

2014-04-01

There are a large number of atomically thin graphitic films with a structure similar to that of graphene. These films have a spread of band gaps relating to their ionicity and, also, to the substrate on which they are grown. Such films could have a range of applications in digital electronics, where graphene is difficult to use. I use the dynamical cluster approximation to show how electron-phonon coupling between film and substrate can enhance these gaps in a way that depends on the range and strength of the coupling. It is found that one of the driving factors in this effect is a charge density wave instability for electrons on a honeycomb lattice that can open a gap in monolayer graphene. The enhancement at intermediate coupling is sufficiently large that spatially varying substrates and superstrates could be used to create heterostructures in thin graphitic films with position-dependent electron-phonon coupling and gaps, leading to advanced electronic components.

Design of a tokamak fusion reactor first wall armor against neutral beam impingement

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

Myers, R.A.

1977-12-01

The maximum temperatures and thermal stresses are calculated for various first wall design proposals, using both analytical solutions and the TRUMP and SAP IV Computer Codes. Beam parameters, such as pulse time, cycle time, and beam power, are varied. It is found that uncooled plates should be adequate for near-term devices, while cooled protection will be necessary for fusion power reactors. Graphite and tungsten are selected for analysis because of their desirable characteristics. Graphite allows for higher heat fluxes compared to tungsten for similar pulse times. Anticipated erosion (due to surface effects) and plasma impurity fraction are estimated. Neutron irradiationmore » damage is also discussed. Neutron irradiation damage (rather than erosion, fatigue, or creep) is estimated to be the lifetime-limiting factor on the lifetime of the component in fusion power reactors. It is found that the use of tungsten in fusion power reactors, when directly exposed to the plasma, will cause serious plasma impurity problems; graphite should not present such an impurity problem.« less

Preliminary burn and impact tests of hybrid polymeric composites. [preventing graphite fiber release

NASA Technical Reports Server (NTRS)

Tompkins, S. S.; Brewer, W. D.

1978-01-01

Free graphite fibers released into the environment from resin matrix composite components, as a result of fire and/or explosion, pose a potential hazard to electrical equipment. An approach to prevent the fibers from becoming airborne is to use hybrid composite materials which retain the fibers at the burn site. Test results are presented for three hybrid composites that were exposed to a simulation of an aircraft fire and explosion. The hybrid systems consisted of 16 plies of graphite-epoxy with two plies of Kevlar-, S-glass-, or boron-epoxy on each face. Two different test environments were used. In one environment, specimens were heated by convection only, and then impacted by a falling mass. In the other environment, specimens were heated by convection and by radiation, but were not impacted. The convective heat flux was about 100-120 kW/m in both environments and the radiative flux was about 110 kW/sq m.

Electrolyte volume effects on electrochemical performance and solid electrolyte interphase in Si-graphite/NMC lithium-ion pouch cells

DOE PAGES

An, Seong Jin; Li, Jianlin; Daniel, Claus; ...

2017-05-15

This study aims to explore the correlations between electrolyte volume, electrochemical performance, and properties of the solid electrolyte interphase in pouch cells with Si-graphite composite anodes. The electrolyte is 1.2 M LiPF 6 in ethylene carbonate:ethylmethyl carbonate with 10 wt.% fluoroethylene carbonate. Single layer pouch cells (100 mAh) were constructed with 15 wt.% Si-graphite/LiNi 0.5Mn 0.3CO 0.2O 2 electrodes. It is found that a minimum electrolyte volume factor of 3.1 times the total pore volume of cell components (cathode, anode, and separator) is needed for better cycling stability. Less electrolyte causes increases in ohmic and charge transfer resistances. Lithium dendritesmore » are observed when the electrolyte volume factor is low. The resistances from the anodes become significant as the cells are discharged. As a result, solid electrolyte interphase thickness grows as the electrolyte volume factor increases and is non-uniform after cycling.« less

Laser induced periodic surface structures formation by nanosecond laser irradiation of poly (ethylene terephthalate) reinforced with Expanded Graphite

NASA Astrophysics Data System (ADS)

Rodríguez-Beltrán, René I.; Hernandez, Margarita; Paszkiewicz, Sandra; Szymczyk, Anna; Rosłaniec, Zbigniew; Ezquerra, Tiberio A.; Castillejo, Marta; Moreno, Pablo; Rebollar, Esther

2018-04-01

We report on the formation of Laser Induced Periodic Surface Structures in poly (ethylene terephthalate) and poly (ethylene terephthalate)/Expanded Graphite films by laser irradiation with nanosecond pulses at 266 nm. The characterization studies show that the quality of the ripples depends strongly on the irradiation time and fluence and the optimal conditions for obtaining LIPSS are affected by the amount of the expanded graphite present in the film due to the differences in crystallinity, thermal conductivity and thermal diffusivity of the nanocomposites. Physicochemical modifications in the materials were inspected by Raman spectroscopy, the colloidal probe technique and contact angle measurements using different liquids. Results show that there is an increase of the hydrophilicity of the surfaces after laser irradiation together with an increase of the surface free energy and in particular of its polar component. Additionally, the adhesion force estimated by the colloidal probe technique increases after laser nanostructuring.

The RaDIATE High-Energy Proton Materials Irradiation Experiment at the Brookhaven Linac Isotope Producer Facility

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

Ammigan, Kavin; et al.

The RaDIATE collaboration (Radiation Damage In Accelerator Target Environments) was founded in 2012 to bring together the high-energy accelerator target and nuclear materials communities to address the challenging issue of radiation damage effects in beam-intercepting materials. Success of current and future high intensity accelerator target facilities requires a fundamental understanding of these effects including measurement of materials property data. Toward this goal, the RaDIATE collaboration organized and carried out a materials irradiation run at the Brookhaven Linac Isotope Producer facility (BLIP). The experiment utilized a 181 MeV proton beam to irradiate several capsules, each containing many candidate material samples formore » various accelerator components. Materials included various grades/alloys of beryllium, graphite, silicon, iridium, titanium, TZM, CuCrZr, and aluminum. Attainable peak damage from an 8-week irradiation run ranges from 0.03 DPA (Be) to 7 DPA (Ir). Helium production is expected to range from 5 appm/DPA (Ir) to 3,000 appm/DPA (Be). The motivation, experimental parameters, as well as the post-irradiation examination plans of this experiment are described.« less

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.

Structural and chemical degradation mechanisms of pure YSZ and its components ZrO2 and Y2O3 in carbon-rich fuel gases.

PubMed

Köck, Eva-Maria; Kogler, Michaela; Götsch, Thomas; Klötzer, Bernhard; Penner, Simon

2016-05-25

Structural and chemical degradation mechanisms of metal-free yttria stabilized zirconia (YSZ-8, 8 mol% Y2O3 in ZrO2) in comparison to its pure oxidic components ZrO2 and Y2O3 have been studied in carbon-rich fuel gases with respect to coking/graphitization and (oxy)carbide formation. By combining operando electrochemical impedance spectroscopy (EIS), operando Fourier-transform infrared (FT-IR) spectroscopy and X-ray photoelectron spectroscopy (XPS), the removal and suppression of CH4- and CO-induced carbon deposits and of those generated in more realistic fuel gas mixtures (syngas, mixtures of CH4 or CO with CO2 and H2O) was examined under SOFC-relevant conditions up to 1273 K and ambient pressures. Surface-near carbidization is a major problem already on the "isolated" (i.e. Nickel-free) cermet components, leading to irreversible changes of the conduction properties. Graphitic carbon deposition takes place already on the "isolated" oxides under sufficiently fuel-rich conditions, most pronounced in the pure gases CH4 and CO, but also significantly in fuel gas mixtures containing H2O and CO2. For YSZ, a comparative quantification of the total amount of deposited carbon in all gases and mixtures is provided and thus yields favorable and detrimental experimental approaches to suppress the carbon formation. In addition, the effectivity and reversibility of removal of the coke/graphite layers was comparably studied in the pure oxidants O2, CO2 and H2O and their effective contribution upon addition to the pure fuel gases CO and CH4 verified.

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.

Evaporation of sessile droplets affected by graphite nanoparticles and binary base fluids.

PubMed

Zhong, Xin; Duan, Fei

2014-11-26

The effects of ethanol component and nanoparticle concentration on evaporation dynamics of graphite-water nanofluid droplets have been studied experimentally. The results show that the formed deposition patterns vary greatly with an increase in ethanol concentration from 0 to 50 vol %. Nanoparticles have been observed to be carried to the droplet surface and form a large piece of aggregate. The volume evaporation rate on average increases as the ethanol concentration increases from 0 to 50 vol % in the binary mixture nanofluid droplets. The evaporation rate at the initial stage is more rapid than that at the late stage to dry, revealing a deviation from a linear fitting line, standing for a constant evaporation rate. The deviation is more intense with a higher ethanol concentration. The ethanol-induced smaller liquid-vapor surface tension leads to higher wettability of the nanofluid droplets. The graphite nanoparticles in ethanol-water droplets reinforce the pinning effect in the drying process, and the droplets with more ethanol demonstrate the depinning behavior only at the late stage. The addition of graphite nanoparticles in water enhances a droplet baseline spreading at the beginning of evaporation, a pinning effect during evaporation, and the evaporation rate. However, with a relatively high nanoparticle concentration, the enhancement is attenuated.

Formation of Reversible Solid Electrolyte Interface on Graphite Surface from Concentrated Electrolytes

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

Lu, Dongping; Tao, Jinhui; Yan, Pengfei

2017-02-10

Interfacial phenomena have always been key determinants for the performance of energy storage technologies. The solid electrolyte interfacial (SEI) layer, pervasive on the surfaces of battery electrodes for numerous chemical couples, directly affects the ion transport, charge transfer and lifespan of the entire energy system. Almost all SEI layers, however, are unstable resulting in the continuous consumption of the electrolyte. Typically, this leads to the accumulation of degradation products on/restructuring of the electrode surface and thus increased cell impedance, which largely limits the long-term operation of the electrochemical reactions. Herein, a completely new SEI formation mechanism has been discovered, inmore » which the electrolyte components reversibly self-assemble into a protective surface coating on a graphite electrode upon changing the potential. In contrast to the established wisdom regarding the necessity of employing the solvent ethylene carbonate (EC) to form a protective SEI layer on graphite, a wide range of EC-free electrolytes are demonstrated for the reversible intercalation/deintercalation of Li+ cations within a graphite lattice, thereby providing tremendous flexibility in electrolyte tailoring for battery couples. This novel finding is broadly applicable and provides guidance for how to control interfacial reactions through the relationship between ion aggregation and solvent decomposition at polarized interfaces.« less

Producibility aspects of advanced composites for an L-1011 Aileron

NASA Technical Reports Server (NTRS)

Van Hamersveld, J.; Fogg, L. D.

1976-01-01

The design of advanced composite aileron suitable for long-term service on transport aircraft includes Kevlar 49 fabric skins on honeycomb sandwich covers, hybrid graphite/Kevlar 49 ribs and spars, and graphite/epoxy fittings. Weight and cost savings of 28 and 20 percent, respectively, are predicted by comparison with the production metallic aileron. The structural integrity of the design has been substantiated by analysis and static tests of subcomponents. The producibility considerations played a key role in the selection of design concepts with potential for low-cost production. Simplicity in fabrication is a major factor in achieving low cost using advanced tooling and manufacturing methods such as net molding to size, draping, forming broadgoods, and cocuring components. A broadgoods dispensing machine capable of handling unidirectional and bidirectional prepreg materials in widths ranging from 12 to 42 inches is used for rapid layup of component kits and covers. Existing large autoclaves, platen presses, and shop facilities are fully exploited.

Quality control of FWC during assembly and commissioning in SST-1 Tokamak

NASA Astrophysics Data System (ADS)

Patel, Hitesh; Santra, Prosenjit; Parekh, Tejas; Biswas, Prabal; Jayswal, Snehal; Chauhan, Pradeep; Paravastu, Yuvakiran; George, Siju; Semwal, Pratibha; Thankey, Prashant; Ramesh, Gattu; Prakash, Arun; Dhanani, Kalpesh; Raval, D. C.; Khan, Ziauddin; Pradhan, Subrata

2017-04-01

First Wall Components (FWC) of SST-1 tokamak, which are in the immediate vicinity of plasma, comprises of limiters, divertors, baffles, passive stabilizers designed to operate long duration (∼1000 s) discharges of elongated plasma. All FWC consist of copper alloy heat sink modules with SS cooling tubes brazed onto it, graphite tiles acting as armour material facing the plasma, and are mounted to the vacuum vessels with suitable Inconel support structures at inter-connected ring & port locations. The FWC are very recently assembled and commissioned successfully inside the vacuum vessel of SST-1 undergoing a rigorous quality control and checks at every stage of the assembly process. This paper will present the quality control aspects and checks of FWC from commencement of assembly procedure, namely material test reports, leak testing of high temperature baked components, assembled dimensional tolerances, leak testing of all welded joints, graphite tile tightening torques, electrical continuity and electrical isolation of passive stabilizers from vacuum vessel, baking and cooling hydraulic connections inside vacuum vessel.

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.

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.

Preparation and Characterization of Graphene Oxide Paper

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

Dikin,D.; Stankovich, S.; Zimney, E.

2007-01-01

Free-standing paper-like or foil-like materials are an integral part of our technological society. Their uses include protective layers, chemical filters, components of electrical batteries or supercapacitors, adhesive layers, electronic or optoelectronic components, and molecular storage. Inorganic 'paper-like' materials based on nanoscale components such as exfoliated vermiculite or mica platelets have been intensively studied and commercialized as protective coatings, high-temperature binders, dielectric barriers and gas-impermeable membranes. Carbon-based flexible graphite foils composed of stacked platelets of expanded graphite have long been used in packing and gasketing applications because of their chemical resistivity against most media, superior sealability over a wide temperature range,more » and impermeability to fluids. The discovery of carbon nanotubes brought about bucky paper, which displays excellent mechanical and electrical properties that make it potentially suitable for fuel cell and structural composite applications. Here we report the preparation and characterization of graphene oxide paper, a free-standing carbon-based membrane material made by flow-directed assembly of individual graphene oxide sheets. This new material outperforms many other paper-like materials in stiffness and strength. Its combination of macroscopic flexibility and stiffness is a result of a unique interlocking-tile arrangement of the nanoscale graphene oxide sheets.« less

Preparation and characterization of graphene oxide paper.

PubMed

Dikin, Dmitriy A; Stankovich, Sasha; Zimney, Eric J; Piner, Richard D; Dommett, Geoffrey H B; Evmenenko, Guennadi; Nguyen, SonBinh T; Ruoff, Rodney S

2007-07-26

Free-standing paper-like or foil-like materials are an integral part of our technological society. Their uses include protective layers, chemical filters, components of electrical batteries or supercapacitors, adhesive layers, electronic or optoelectronic components, and molecular storage. Inorganic 'paper-like' materials based on nanoscale components such as exfoliated vermiculite or mica platelets have been intensively studied and commercialized as protective coatings, high-temperature binders, dielectric barriers and gas-impermeable membranes. Carbon-based flexible graphite foils composed of stacked platelets of expanded graphite have long been used in packing and gasketing applications because of their chemical resistivity against most media, superior sealability over a wide temperature range, and impermeability to fluids. The discovery of carbon nanotubes brought about bucky paper, which displays excellent mechanical and electrical properties that make it potentially suitable for fuel cell and structural composite applications. Here we report the preparation and characterization of graphene oxide paper, a free-standing carbon-based membrane material made by flow-directed assembly of individual graphene oxide sheets. This new material outperforms many other paper-like materials in stiffness and strength. Its combination of macroscopic flexibility and stiffness is a result of a unique interlocking-tile arrangement of the nanoscale graphene oxide sheets.

BRAZING ALLOYS

DOEpatents

Donnelly, R.G.; Gilliland, R.G.; Slaughter, G.M.

1963-02-26

A brazing alloy which, in the molten state, is characterized by excellent wettability and flowability, said alloy being capable of forming a corrosion resistant brazed joint wherein at least one component of said joint is graphite and the other component is a corrosion resistant refractory metal, said alloy consisting essentially of 20 to 50 per cent by weight of gold, 20 to 50 per cent by weight of nickel, and 15 to 45 per cent by weight of molybdenum. (AEC)

Non-precious metal catalysts prepared from precursor comprising cyanamide

DOEpatents

Chung, Hoon Taek; Zelenay, Piotr

2015-10-27

Catalyst comprising graphitic carbon and methods of making thereof; said graphitic carbon comprising a metal species, a nitrogen-containing species and a sulfur containing species. A catalyst for oxygen reduction reaction for an alkaline fuel cell was prepared by heating a mixture of cyanamide, carbon black, and a salt selected from an iron sulfate salt and an iron acetate salt at a temperature of from about 700.degree. C. to about 1100.degree. C. under an inert atmosphere. Afterward, the mixture was treated with sulfuric acid at elevated temperature to remove acid soluble components, and the resultant mixture was heated again under an inert atmosphere at the same temperature as the first heat treatment step.

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

Self repairing composites for drone air vehicles

NASA Astrophysics Data System (ADS)

Dry, Carolyn

2015-04-01

The objective of this effort was to demonstrate the feasibility of impact-initiated delivery of repair chemicals through hollow fiber architectures embedded within graphite fiber reinforced polymer matrix composites, representative of advanced drone aircraft component material systems. Self-repairing structures through coupon and elements were demonstrated, and evaluated.

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.

Modelling the graphite fracture mechanisms

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

Jacquemoud, C.; Marie, S.; Nedelec, M.

2012-07-01

In order to define a design criterion for graphite components, it is important to identify the physical phenomena responsible for the graphite fracture, to include them in a more effective modelling. In a first step, a large panel of experiments have been realised in order to build up an important database; results of tensile tests, 3 and 4 point bending tests on smooth and notched specimens have been analysed and have demonstrated an important geometry related effects on the behavior up to fracture. Then, first simulations with an elastic or an elastoplastic bilinear constitutive law have not made it possiblemore » to simulate the experimental fracture stress variations with the specimen geometry, the fracture mechanisms of the graphite being at the microstructural scale. That is the reason why a specific F.E. model of the graphite structure has been developed in which every graphite grain has been meshed independently, the crack initiation along the basal plane of the particles as well as the crack propagation and coalescence have been modelled too. This specific model has been used to test two different approaches for fracture initiation: a critical stress criterion and two criteria of fracture mechanic type. They are all based on crystallographic considerations as a global critical stress criterion gave unsatisfactory results. The criteria of fracture mechanic type being extremely unstable and unable to represent the graphite global behaviour up to the final collapse, the critical stress criterion has been preferred to predict the results of the large range of available experiments, on both smooth and notched specimens. In so doing, the experimental observations have been correctly simulated: the geometry related effects on the experimental fracture stress dispersion, the specimen volume effects on the macroscopic fracture stress and the crack propagation at a constant stress intensity factor. In addition, the parameters of the criterion have been related to experimental observations: the local crack initiation stress of 8 MPa corresponds to the non-linearity apparition on the global behavior observed experimentally and the the maximal critical stress defined for the particle of 30 MPa is equivalent to the fracture stress of notched specimens. This innovative combination of crack modelling and a local crystallographic critical stress criterion made it possible to understand that cleavage initiation and propagation in the graphite microstructure was driven by a mean critical stress criterion. (authors)« less

VERY LARGE INTERSTELLAR GRAINS AS EVIDENCED BY THE MID-INFRARED EXTINCTION

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

Wang, Shu; Jiang, B. W.; Li, Aigen, E-mail: shuwang@mail.bnu.edu.cn, E-mail: bjiang@bnu.edu.cn, E-mail: wanshu@missouri.edu, E-mail: lia@missouri.edu

The sizes of interstellar grains are widely distributed, ranging from a few angstroms to a few micrometers. The ultraviolet (UV) and optical extinction constrains the dust in the size range of a couple hundredths of micrometers to several submicrometers. The near and mid infrared (IR) emission constrains the nanometer-sized grains and angstrom-sized very large molecules. However, the quantity and size distribution of micrometer-sized grains remain unknown because they are gray in the UV/optical extinction and they are too cold and emit too little in the IR to be detected by IRAS, Spitzer, or Herschel. In this work, we employ themore » ∼3–8 μm mid-IR extinction, which is flat in both diffuse and dense regions to constrain the quantity, size, and composition of the μm-sized grain component. We find that, together with nano- and submicron-sized silicate and graphite (as well as polycyclic aromatic hydrocarbons), μm-sized graphite grains with C/H ≈ 137 ppm and a mean size of ∼1.2 μm closely fit the observed interstellar extinction of the Galactic diffuse interstellar medium from the far-UV to the mid-IR, as well as the near-IR to millimeter thermal emission obtained by COBE/DIRBE, COBE/FIRAS, and Planck up to λ ≲ 1000 μm. The μm-sized graphite component accounts for ∼14.6% of the total dust mass and ∼2.5% of the total IR emission.« less

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

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

Harp, Jason Michael; Stempien, John Dennis; Demkowicz, Paul Andrew

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO 2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. Thesemore » data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO 2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO 2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.« less

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

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

Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These datamore » were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.« less

Unusually conductive carbon-inherently conducting polymer (ICP) composites: Synthesis and characterization

NASA Astrophysics Data System (ADS)

Bourdo, Shawn Edward

Two groups of materials that have recently come to the forefront of research initiatives are carbon allotropes, especially nanotubes, and conducting polymers-more specifically inherently conducting polymers. The terms conducting polymers and inherently conducting polymers sometimes are used interchangeably without fully acknowledging a major difference in these terms. Conducting polymers (CPs) and inherently conducting polymers (ICPs) are both polymeric materials that conduct electricity, but the difference lies in how each of these materials conducts electricity. For CPs of the past, an electrically conductive filler such as metal particles, carbon black, or graphite would be blended into a polymer (insulator) allowing for the CP to carry an electric current. An ICP conducts electricity due to the intrinsic nature of its chemical structure. The two materials at the center of this research are graphite and polyaniline. For the first time, a composite between carbon allotropes (graphite) and an inherently conducting polymer (PANI) has exhibited an electrical conductivity greater than either of the two components. Both components have a plethora of potential applications and therefore the further investigation could lead to use of these composites in any number of technologies. Touted applications that use either conductive carbons or ICPs exist in a wide range of fields, including electromagnetic interference (EMI) shielding, radar evasion, low power rechargeable batteries, electrostatic dissipation (ESD) for anti-static textiles, electronic devices, light emitting diodes (LEDs), corrosion prevention, gas sensors, super capacitors, photovoltaic cells, and resistive heating. The main motivation for this research has been to investigate the connection between an observed increase in conductivity and structure of composites. Two main findings have resulted from the research as related to the observed increase in conductivity. The first was the structural evidence from Raman spectroscopy, X-ray diffraction, and thermal analysis suggesting a more crystalline graphite matrix due to intimate interactions with PANI that resulted in a charge transfer. Confirmation of charge transfer was observed through magnetic susceptibility, electron paramagnetic resonance, and temperature dependent electrical conductivity studies.

Evaluation of a metal fuselage panel selectively reinforced with filamentary composites for space shuttle application

NASA Technical Reports Server (NTRS)

Wennhold, W. F.

1974-01-01

The use of high strength and modulus of advanced filamentary composites to reduce the structural weight of aerospace vehicles was investigated. Application of the technology to space shuttle components was the primary consideration. The mechanical properties for the boron/epoxy, graphite/epoxy, and polyimide data are presented. Structural testing of two compression panel components was conducted in a simulated space shuttle thermal environment. Results of the tests are analyzed.

Coated metal sintering carriers for fuel cell electrodes

DOEpatents

Donelson, Richard; Bryson, E. S.

1998-01-01

A carrier for conveying components of a fuel cell to be sintered through a sintering furnace. The carrier comprises a metal sheet coated with a water-based carbon paint, the water-based carbon paint comprising water, powdered graphite, an organic binder, a wetting agent, a dispersing agent and a defoaming agent.

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

Visible-light-enhanced interactions of hydrogen sulfide with composites of zinc (oxy)hydroxide with graphite oxide and graphene.

PubMed

Seredych, Mykola; Mabayoje, Oluwaniyi; Bandosz, Teresa J

2012-01-17

Composites of zinc(oxy)hydroxide-graphite oxide and of zinc(oxy)hydroxide-graphene were used as adsorbents of hydrogen sulfide under ambient conditions. The initial and exhausted samples were characterized by XRD, FTIR, potentiometric titration, EDX, thermal analysis, and nitrogen adsorption. An increase in the amount of H(2)S adsorbed/oxidized on their surfaces in comparison with that of pure Zn(OH)(2) is linked to the structure of the composite, the relative number of terminal hydroxyls, and the kind of graphene-based phase used. Although terminal groups are activated by a photochemical process, the graphite oxide component owing to the chemical bonds with the zinc(oxy)hydroxide phase and conductive properties helps in electron transfer, leading to more efficient oxygen activation via the formation of superoxide ions. Elemental sulfur, zinc sulfide, sulfite, and sulfate are formed on the surface. The formation of sulfur compounds on the surface of zinc(oxy)hydroxide during the course of the breakthrough experiments and thus Zn(OH)(2)-ZnS heterojunctions can also contribute to the increased surface activity of our materials. The results show the superiority of graphite oxide in the formation of composites owing to its active surface chemistry and the possibility of interface bond formation, leading to an increase in the number of electron-transfer reactions. © 2011 American Chemical Society

Enhancing the oxidation resistance of graphite by applying an SiC coat with crack healing at an elevated temperature

NASA Astrophysics Data System (ADS)

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

2016-08-01

The potential of reducing the oxidation of the supporting graphite components during normal and/or accident conditions in the Very High Temperature Reactor (VHTR) design has been studied. In this work efforts have been made to slow the oxidation process of the graphite with a thin SiC coating (∼ 10 μm). Upon heating at ≥ 1173 K in air, the spallations and cracks were formed in the dense columnar structured SiC coating layer grown on the graphite with a functionally gradient electron beam physical vapor deposition (EB-PVD. In accordance with the formations of these defects, the sample was vigorously oxidized, leaving only the SiC coating layer. Then, efforts were made to heal the surface defects using additional EB-PVD with ion beam bombardment and chemical vapor deposition (CVD). The EB-PVD did not effectively heal the cracks. But, the CVD was more appropriate for crack healing, likely due to its excellent crack line filling capability with a high density and high aspect ratio. It took ∼ 34 min for the 20% weight loss of the CVD crack healed sample in the oxidation test with annealing at 1173 K, while it took ∼ 8 min for the EB-PVD coated sample, which means it took ∼4 times longer at 1173 K for the same weight reduction in this experimental set-up.

Solid sampling determination of magnesium in lithium niobate crystals by graphite furnace atomic absorption spectrometry

NASA Astrophysics Data System (ADS)

Dravecz, Gabriella; Laczai, Nikoletta; Hajdara, Ivett; Bencs, László

2016-12-01

The vaporization/atomization processes of Mg in high-resolution continuum source graphite furnace atomic absorption spectrometry (HR-CS-GFAAS) were investigated by evaporating solid (powder) samples of lithium niobate (LiNbO3) optical single crystals doped with various amounts of Mg in a transversally heated graphite atomizer (THGA). Optimal analytical conditions were attained by using the Mg I 215.4353 nm secondary spectral line. An optimal pyrolysis temperature of 1500 °C was found for Mg, while the compromise atomization temperature in THGAs (2400 °C) was applied for analyte vaporization. The calibration was performed against solid (powered) lithium niobate crystal standards. The standards were prepared with exactly known Mg content via solid state fusion of the oxide components of the matrix and analyte. The correlation coefficient (R value) of the linear calibration was not worse than 0.9992. The calibration curves were linear in the dopant concentration range of interest (0.74-7.25 mg/g Mg), when dosing 3-10 mg of the powder samples into the graphite sample insertion boats. The Mg content of the studied 19 samples was in the range of 1.69-4.13 mg/g. The precision of the method was better than 6.3%. The accuracy of the results was verified by means of flame atomic absorption spectrometry with solution sample introduction after digestion of several crystal samples.

Development of a Continuum Damage Mechanics Material Model of a Graphite-Kevlar(Registered Trademark) Hybrid Fabric for Simulating the Impact Response of Energy Absorbing Kevlar(Registered Trademark) Hybrid Fabric for Simulating the Impact Response of Energy Absorbing

NASA Technical Reports Server (NTRS)

Jackson, Karen E.; Fasanella, Edwin L.; Littell, Justin D.

2017-01-01

This paper describes the development of input properties for a continuum damage mechanics based material model, Mat 58, within LS-DYNA(Registered Trademark) to simulate the response of a graphite-Kevlar(Registered Trademark) hybrid plain weave fabric. A limited set of material characterization tests were performed on the hybrid graphite-Kevlar(Registered Trademark) fabric. Simple finite element models were executed in LS-DYNA(Registered Trademark) to simulate the material characterization tests and to verify the Mat 58 material model. Once verified, the Mat 58 model was used in finite element models of two composite energy absorbers: a conical-shaped design, designated the "conusoid," fabricated of four layers of hybrid graphite-Kevlar(Registered Trademark) fabric; and, a sinusoidal-shaped foam sandwich design, designated the "sinusoid," fabricated of the same hybrid fabric face sheets with a foam core. Dynamic crush tests were performed on components of the two energy absorbers, which were designed to limit average vertical accelerations to 25- to 40-g, to minimize peak crush loads, and to generate relatively long crush stroke values under dynamic loading conditions. Finite element models of the two energy absorbers utilized the Mat 58 model that had been verified through material characterization testing. Excellent predictions of the dynamic crushing response were obtained.

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2012 CFR

2012-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2013 CFR

2013-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2014 CFR

2014-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

Evaluation of flawed composite structural components under static and cyclic loading. [fatigue life of graphite-epoxy composite materials

NASA Technical Reports Server (NTRS)

Porter, T. R.

1979-01-01

The effects of initial defects on the fatigue and fracture response of graphite-epoxy composite laminates are presented. The structural laminates investigated were a typical angle ply laminate, a polar/hoop wound pressure vessel laminate, and a typical engine fan blade laminate. Defects investigated were full and half penetration circular holes, full and half penetration slits, and countersink holes. The effects of the defect size and type on the static fracture strength, fatigue performance, and residual static strength are shown as well as the results of loadings on damage propagation in composite laminates. The data obtained were used to define proof test levels as a qualification procedure in composite structure subjected to cyclic loading.

Hindered Glymes for Graphite-Compatible Electrolytes.

PubMed

Shanmukaraj, Devaraj; Grugeon, Sylvie; Laruelle, Stephane; Armand, Michel

2015-08-24

Organic carbonate mixtures are used almost exclusively as lithium battery electrolyte solvents. The linear compounds (dimethyl carbonate, diethyl carbonate, ethyl methyl carbonate) act mainly as thinner for the more viscous and high-melting ethylene carbonate but are the least stable component and have low flash points; these are serious handicaps for lifetime and safety. Polyethers (glymes) are useful co-solvents, but all formerly known representatives solvate Li(+) strongly enough to co-intercalate in the graphite negative electrode and exfoliate it. We have put forward a new electrolyte composition comprising a polyether to which a bulky tert-butyl group is attached ("hindered glyme"), thus completely preventing co-intercalation while maintaining good conductivity. This alkyl-carbonate-free electrolyte shows remarkable cycle efficiency of the graphite electrode, not only at room temperature, but also at 50 and 70 °C in the presence of lithium bis(fluorosulfonimide). The two-ethylene-bridge hindered glyme has a high boiling point and a flash point of 80 °C, a considerable advantage for safety. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Study to investigate design, fabrication and test of low cost concepts for large hybrid composite helicopter fuselage, phase 1

NASA Technical Reports Server (NTRS)

Adams, K. M.; Lucas, J. J.

1975-01-01

The development of a frame/stringer/skin fabrication technique for composite airframe construction was studied as a low cost approach to the manufacture of large helicopter airframe components. A center cabin aluminum airframe section of the Sikorsky CH-53D helicopter was selected for evaluation as a composite structure. The design, as developed, is composed of a woven KEVLAR-49/epoxy skin and graphite/epoxy frames and stringers. To support the selection of this specific design concept a materials study was conducted to develop and select a cure compatible graphite and KEVLAR-49/epoxy resin system, and a foam system capable of maintaining shape and integrity under the processing conditions established. The materials selected were, Narmco 5209/Thornel T-300 graphite, Narmco 5209/KEVLAR-49 woven fabric, and Stathane 8747 polyurethane foam. Eight specimens were fabricated, representative of the frame, stringer, and splice joint attachments. Evaluation of the results of analysis and test indicate that design predictions are good to excellent except for some conservatism of the complex frame splice.

Residual-strength tests of L-1011 vertical fin components after 10 and 20 years of simulated flight service

NASA Technical Reports Server (NTRS)

Lopez, O. F.

1984-01-01

Part of the NASA/ACEE Program was to determine the effect of long-term durability testing on the residual strength of graphite-epoxy cover panel and spar components of the Lockheed L-1011 aircraft vertical stabilizer. The results of these residual strength tests are presented herein. The structural behavior and failure mode of both cover panel and spar components were addressed, and the test results obtained were compared with the static test results generated by Lockheed. The effect of damage on one of the spar specimens was described.

Advanced technology for extended endurance alkaline fuel cells

NASA Technical Reports Server (NTRS)

Sheibley, D. W.; Martin, R. A.

1987-01-01

Advanced components have been developed for alkaline fuel cells with a view to the satisfaction of NASA Space Station design requirements for extended endurance. The components include a platinum-on-carbon catalyst anode, a potassium titanate-bonded electrolyte matrix, a lightweight graphite electrolyte reservoir plate, a gold-plated nickel-perforated foil electrode substrate, a polyphenylene sulfide cell edge frame material, and a nonmagnesium cooler concept. When incorporated into the alkaline fuel cell unit, these components are expected to yield regenerative operation in a low earth orbit Space Station with a design life greater than 5 years.

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.

Development and test of advanced composite components. Center Directors discretionary fund program

NASA Technical Reports Server (NTRS)

Faile, G.; Hollis, R.; Ledbetter, F.; Maldonado, J.; Sledd, J.; Stuckey, J.; Waggoner, G.; Engler, E.

1985-01-01

This report describes the design, analysis, fabrication, and test of a complex bathtub fitting. Graphite fibers in an epoxy matrix were utilized in manufacturing of 11 components representing four different design and layup concepts. Design allowables were developed for use in the final stress analysis. Strain gage measurements were taken throughout the static load test and correlation of test and analysis data were performed, yielding good understanding of the material behavior and instrumentation requirements for future applications.

Advanced rotary engine components utilizing fiber reinforced Mg castings

NASA Technical Reports Server (NTRS)

Goddard, D.; Whitman, W.; Pumphrey, R.; Lee, C.-M.

1986-01-01

Under a two-phase program sponsored by NASA, the technology for producing advanced rotary engine components utilizing graphite fiber-reinforced magnesium alloy casting is being developed. In Phase I, the successful casting of a simulated intermediate housing was demonstrated. In Phase II, the goal is to produce an operating rotor housing. The effort involves generation of a material property data base, optimization of parameters, and development of wear- and corrosion-resistant cast surfaces and surface coatings. Results to date are described.

Assembly & Metrology of First Wall Components of SST-1

NASA Astrophysics Data System (ADS)

Parekh, Tejas; Santra, Prosenjit; Biswas, Prabal; Patel, Hiteshkumar; Paravastu, Yuvakiran; Jaiswal, Snehal; Chauhan, Pradeep; Babu, Gattu Ramesh; A, Arun Prakash; Bhavsar, Dhaval; Raval, Dilip C.; Khan, Ziauddin; Pradhan, Subrata

2017-04-01

First Wall Components (FWC) of SST-1 tokamak, which are in the immediate vicinity of plasma comprises of limiters, divertors, baffles, passive stabilizers are designed to operate long duration (1000 s) discharges of elongated plasma. All FWC consists of a copper alloy heat sink modules with SS cooling tubes brazed onto it, graphite tiles acting as armour material facing the plasma, and are mounted to the vacuum vessels with suitable Inconel support structures at ring & port locations. The FWC are very recently assembled and commissioned successfully inside the vacuum vessel of SST-1 undergoing a meticulous planning of assembly sequence, quality checks at every stage of the assembly process. This paper will present the metrology aspects & procedure of each FWC, both outside the vacuum vessel, and inside the vessel, assembly tolerances, tools, equipment and jig/fixtures, used at each stage of assembly, starting from location of support bases on vessel rings, fixing of copper modules on support structures, around 3800 graphite tile mounting on 136 copper modules with proper tightening torques, till final toroidal and poloidal geometry of the in-vessel components are obtained within acceptable limits, also ensuring electrical continuity of passive stabilizers to form a closed saddle loop, electrical isolation of passive stabilizers from vacuum vessel.

Impact Testing and Simulation of Composite Airframe Structures

NASA Technical Reports Server (NTRS)

Jackson, Karen E.; Littell, Justin D.; Horta, Lucas G.; Annett, Martin S.; Fasanella, Edwin L.; Seal, Michael D., II

2014-01-01

Dynamic tests were performed at NASA Langley Research Center on composite airframe structural components of increasing complexity to evaluate their energy absorption behavior when subjected to impact loading. A second objective was to assess the capabilities of predicting the dynamic response of composite airframe structures, including damage initiation and progression, using a state-of-the-art nonlinear, explicit transient dynamic finite element code, LS-DYNA. The test specimens were extracted from a previously tested composite prototype fuselage section developed and manufactured by Sikorsky Aircraft Corporation under the US Army's Survivable Affordable Repairable Airframe Program (SARAP). Laminate characterization testing was conducted in tension and compression. In addition, dynamic impact tests were performed on several components, including I-beams, T-sections, and cruciform sections. Finally, tests were conducted on two full-scale components including a subfloor section and a framed fuselage section. These tests included a modal vibration and longitudinal impact test of the subfloor section and a quasi-static, modal vibration, and vertical drop test of the framed fuselage section. Most of the test articles were manufactured of graphite unidirectional tape composite with a thermoplastic resin system. However, the framed fuselage section was constructed primarily of a plain weave graphite fabric material with a thermoset resin system. Test data were collected from instrumentation such as accelerometers and strain gages and from full-field photogrammetry.

Coated metal sintering carriers for fuel cell electrodes

DOEpatents

Donelson, R.; Bryson, E.S.

1998-11-10

A carrier is described for conveying components of a fuel cell to be sintered through a sintering furnace. The carrier comprises a metal sheet coated with a water-based carbon paint, the water-based carbon paint comprising water, powdered graphite, an organic binder, a wetting agent, a dispersing agent and a defoaming agent.

Some observations on the fracture of austempered ductile iron

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

Fan, Z.K.; Smallman, R.E.

1994-07-15

There is extensive work on the fracture of steel with ferrite or/and austenite structure, but little on crack propagation in austempered ductile iron (ADI) whose microstructure also comprises austenite and ferrite (in the form of bainitic ferrite) but with graphite nodules in the matrix. Because of its good combination of wear resistance and toughness, and its low density and low cost (compared with forge steel), ADI has been widely used for various kinds of engineering components, such as gears, crankshafts, vehicle components, sprockets, cutting and digging tools etc. The matrix of ADI can withstand a certain amount of deformation beforemore » fracture during tensile or impact testing; for example, the elongation of ADI (grade 1050/700/7 to ASTM Standard) can reach 7--10% during tensile testing. However, the graphite nodules in the matrix cannot deform and hence are barriers to matrix deformation and give rise to crack initiation. In addition, carbides may precipitate in the bainitic ferrite laths or at the ferrite/austenite interfaces and these may also influence the fracture of ADI and produce characteristic features.« less

Vacuum Plasma Spray Forming of Tungsten Lorentz Force Accelerator Components

NASA Technical Reports Server (NTRS)

Zimmerman, Frank R.

2004-01-01

The Vacuum Plasma Spray (VPS) Laboratory at NASA's Marshall Space Flight Center, working with the Jet Propulsion Laboratory, has developed and demonstrated a fabrication technique using the VPS process to form anode and cathode sections for a Lorentz force accelerator made from tungsten. Lorentz force accelerators are an attractive form of electric propulsion that provides continuous, high-efficiency propulsion at useful power levels for such applications as orbit transfers or deep space missions. The VPS process is used to deposit refractory metals such as tungsten onto a graphite mandrel of the desired shape. Because tungsten is reactive at high temperatures, it is thermally sprayed in an inert environment where the plasma gun melts and deposits the molten metal powder onto a mandrel. A three-axis robot inside the chamber controls the motion of the plasma spray torch. A graphite mandrel acts as a male mold, forming the required contour and dimensions for the inside surface of the anode or cathode of the accelerator. This paper describes the processing techniques, design considerations, and process development associated with the VPS forming of Lorentz force accelerator components.

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

A Detailed FLUKA-2005 Monte Carlo Simulation for the ATIC Detector

NASA Technical Reports Server (NTRS)

Gunasingha, R. M.; Fazely, A. R.; Adams, J. H.; Ahn, H. S.; Bashindzhagyan, G. L.; Batkov, K. E.; Chang, J.; Christl, M.; Ganel, O.; Guzik, T. G.

2006-01-01

We have performed a detailed Monte Carlo (MC) calculation for the Advanced thin Ionization Calorimeter (ATIC) detector using the MC code FLUKA-2005 which is capable of simulating particles up to 10 PeV. The ATIC detector has completed two successful balloon flights from McMurdo, Antarctica lasting a total of more than 35 days. ATIC is designed as a multiple, long duration balloon Bight, investigation of the cosmic ray spectra from below 50 GeV to near 100 TeV total energy; using a fully active Bismuth Germanate @GO) calorimeter. It is equipped with a large mosaic of silicon detector pixels capable of charge identification and as a particle tracking system, three projective layers of x-y scintillator hodoscopes were employed, above, in the middle and below a 0.75 nuclear interaction length graphite target. Our calculations are part of an analysis package of both A- and energy-dependences of different nuclei interacting with the ATIC detector. The MC simulates the responses of different components of the detector such as the Simatrix, the scintillator hodoscopes and the BGO calorimeter to various nuclei. We also show comparisons of the FLUKA-2005 MC calculations with a GEANT calculation and data for protons, He and CNO.

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

Advanced manufacturing development of a composite empennage component for L-1011 aircraft. Phase 2: Design and analysis

NASA Technical Reports Server (NTRS)

Jackson, A. C.; Crocker, J. F.; Ekvall, J. C.; Eudaily, R. R.; Mosesian, B.; Vancleave, R. R.; Vanhamersveld, J.

1981-01-01

The composite fin design consists of two one-piece cocured covers, two one-piece cocured spars and eleven ribs. The lower ribs are truss ribs with graphite/epoxy caps and aluminum truss members. The upper three ribs are a sandwich design with graphite/epoxy face sheets and a syntactic epoxy core. The design achieves a 27% weight saving compared to the metal box. The fastener count has been reduced from over 40,000 to less than 7000. The structural integrity of the composite fin was verified by analysis and test. The static, fail-safe and flutter analyses were completed. An extensive test program has established the material behavior under a range of conditions and critical subcomponents were tested to verify the structural concepts.

Strain and dynamic measurements using fiber optic sensors embedded into graphite/epoxy tubes

NASA Technical Reports Server (NTRS)

Dehart, D. W.; Doederlein, T.; Koury, J.; Rogowski, R. S.; Heyman, J. S.; Holben, M. S., Jr.

1989-01-01

Graphite/epoxy tubes were fabricated with embedded optical fibers to evaluate the feasibility of monitoring strains with a fiber optic technique. Resistance strain gauges were attached to the tubes to measure strain at four locations along the tube for comparison with the fiber optic sensors. Both static and dynamic strain measurements were made with excellent agreement between the embedded fiber optic strain sensor and the strain gauges. Strain measurements of 10(exp -7) can be detected with the optical phase locked loop (OPLL) system using optical fiber. Because of their light weight, compatibility with composites, immunity to electromagnetic interference, and based on the static and dynamic results obtained, fiber optic sensors embedded in composites may be useful as the sensing component of smart structures.

Three-dimensional direct laser written graphitic electrical contacts to randomly distributed components

NASA Astrophysics Data System (ADS)

Dorin, Bryce; Parkinson, Patrick; Scully, Patricia

2018-04-01

The development of cost-effective electrical packaging for randomly distributed micro/nano-scale devices is a widely recognized challenge for fabrication technologies. Three-dimensional direct laser writing (DLW) has been proposed as a solution to this challenge, and has enabled the creation of rapid and low resistance graphitic wires within commercial polyimide substrates. In this work, we utilize the DLW technique to electrically contact three fully encapsulated and randomly positioned light-emitting diodes (LEDs) in a one-step process. The resolution of the contacts is in the order of 20 μ m, with an average circuit resistance of 29 ± 18 kΩ per LED contacted. The speed and simplicity of this technique is promising to meet the needs of future microelectronics and device packaging.

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.

Upshot of natural graphite inclusion on the performance of porous conducting carbon fiber paper in a polymer electrolyte membrane fuel cell

NASA Astrophysics Data System (ADS)

Kaushal, Shweta; Negi, Praveen; Sahu, A. K.; Dhakate, S. R.

2017-09-01

Porous conducting carbon fiber paper (PCCFP) is one of the vital component of the gas diffusion layer (GDL) in a fuel cell. This PCCFP serves as the most suitable substrate for the GDL due to its electrical conductivity, mechanical properties, and porosity. In this approach, carbon fiber composite papers were developed by incorporating different fractions of natural graphite (NG) in the matrix phase, i.e. Phenolic resin, and using the combined process of paper making and carbon-carbon composite formation technique. These prepared samples were then heat treated at 1800 °C in an inert atmosphere. The effect of natural graphite incorporation was ascertained by characterizing porous carbon paper by various techniques i.e. X-ray diffraction, Raman spectroscopy, Scanning electron microscopy, electrical and mechanical properties, and I-V performance in a unit fuel cell assembly. The inclusion of NG certainly enhance the properties of the carbon matrix as well as improving the conductive path of carbon fibers. In this study addition of 1 wt.% of natural graphite demonstrated a significant improvement in the electrical conductivity and performance of PCCFP and resulted in the improvement of power density from 361-563 mW cm-2. This paper reports that the uniform dispersion of NG was able to generate a maximum number of macrosize pores in the carbon paper that strengthened the flexural modulus from 4 to 12 GPa without compromising the porosity required for the GDL.

X-33 LH2 Tank Failure Investigation Findings

NASA Technical Reports Server (NTRS)

Niedermeyer, Mindy; Clinton, R. G., Jr. (Technical Monitor)

2000-01-01

This presentation focuses on the tank history, test objectives, failure description, investigation and conclusions. The test objectives include verify structural integrity at 105% expected flight load limit varying the following parameters: cryogenic temperature; internal pressure; and mechanical loading. The Failure description includes structural component of the aft body, quad-lobe design, and sandwich - honeycomb graphite epoxy construction.

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.

Castellated tiles as the beam-facing components for the diagnostic calorimeter of the negative ion source SPIDER

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

Peruzzo, S., E-mail: simone.peruzzo@igi.cnr.it; Cervaro, V.; Dalla Palma, M.

2016-02-15

This paper presents the results of numerical simulations and experimental tests carried out to assess the feasibility and suitability of graphite castellated tiles as beam-facing component in the diagnostic calorimeter of the negative ion source SPIDER (Source for Production of Ions of Deuterium Extracted from Radio frequency plasma). The results indicate that this concept could be a reliable, although less performing, alternative for the present design based on carbon fiber composite tiles, as it provides thermal measurements on the required spatial scale.

Castellated tiles as the beam-facing components for the diagnostic calorimeter of the negative ion source SPIDER

NASA Astrophysics Data System (ADS)

Peruzzo, S.; Cervaro, V.; Dalla Palma, M.; Delogu, R.; De Muri, M.; Fasolo, D.; Franchin, L.; Pasqualotto, R.; Pimazzoni, A.; Rizzolo, A.; Tollin, M.; Zampieri, L.; Serianni, G.

2016-02-01

This paper presents the results of numerical simulations and experimental tests carried out to assess the feasibility and suitability of graphite castellated tiles as beam-facing component in the diagnostic calorimeter of the negative ion source SPIDER (Source for Production of Ions of Deuterium Extracted from Radio frequency plasma). The results indicate that this concept could be a reliable, although less performing, alternative for the present design based on carbon fiber composite tiles, as it provides thermal measurements on the required spatial scale.

Finite Element Simulation of Solid Rocket Booster Separation Motors During Motor Firing

NASA Technical Reports Server (NTRS)

Yu. Weiping; Crane, Debora J.

2007-01-01

One of the toughest challenges facing Solid Rocket Booster (SRB) engineers is to ensure that any design changes made to the Shuttle-Derived Booster Separation Motors (BSM) for future space exploration vehicles is able to withstand the increasingly hostile motor firing environment without cracking its critical component - the graphite throat. This paper presents a critical analysis methodology and techniques for assessing effects of BSM design changes with great accuracy and precision. For current Space Shuttle operation, the motor firing occurs at SRB separation - approximately 125 seconds after Shuttle launch at an altitude of about 28 miles. The motor operation event lasts about two seconds, however, the surface temperature of the graphite throat increases approximately 3400 F in less than one second with a corresponding increase in surface pressure of approximately 2200 pounds per square inch (psi) in less than one-tenth of a second. To capture this process fully and accurately, a two-phase sequentially coupled thermal-mechanical finite element approach was developed. This method allows the time- and location-dependent pressure fields to interact with the spatial-temporal thermal fields throughout the operation. The material properties of graphite throat are orthotropic and temperature-dependent. The analysis involves preload and multiple body contacts.

Lithium dendrite and solid electrolyte interphase investigation using OsO4

NASA Astrophysics Data System (ADS)

Zier, Martin; Scheiba, Frieder; Oswald, Steffen; Thomas, Jürgen; Goers, Dietrich; Scherer, Torsten; Klose, Markus; Ehrenberg, Helmut; Eckert, Jürgen

2014-11-01

Osmium tetroxide (OsO4) staining, commonly used to enhance scattering contrast in electron microscopy of biologic tissue and polymer blends, has been adopted for studies of graphite anodes in lithium-ion batteries. OsO4 shows a coordinated reaction with components of the solid electrolyte interphase (SEI) and lithium dendrites, thereby increasing material contrast for scanning electron microscopy investigations. Utilizing the high affinity of lithium metal to react with osmium tetroxide it was possible to localize even small lithium deposits on graphite electrodes. In spite of their reaction with the OsO4 fume, the lithium dendrite morphology remains almost untouched by the staining procedure, offering information on the dendrite growth process. Correlating the quantity of osmium detected with the amount of residual ("dead") lithium of a discharged electrode, it was possible to obtain a practical measure for lithium plating and stripping efficiencies. EDX mappings allowed for a localization of electrochemically stripped lithium dendrites by their residual stained SEI shells. Cross sections, prepared by focused ion beam (FIB) of cycled graphite electrodes treated with OsO4, revealed important information about deposition and distribution of metallic lithium and the electrolyte reduction layer across the electrode.

Brazed graphite/refractory metal composites for first-wall protection elements

NASA Astrophysics Data System (ADS)

Šmid, I.; Croessmann, C. D.; Salmonson, J. C.; Whitley, J. B.; Kny, E.; Reheis, N.; Kneringer, G.; Nickel, H.

1991-03-01

The peak surface heat flux deposition on divertor elements of near term fusion devices is expected to exceed 10 MW/m 2. The needed reliability of brazed plasma interactive components, particularly under abnormal operating conditions with peak surface temperatures well beyond 1000°C, makes refractory metallic substrates and brazes with a high melting point very attractive. TZM, a high temperature alloy of molybdenum, and isotropic graphite, materials very closely matched in their thermal expansion, were brazed with four high-temperature brazes. The brazes used were Zr, 90Ni/10Ti, 90Cu/10Ti and 70Ag/27Cu/3Ti (nominal composition prior to brazing, wt%). The resulting composite tiles of 50 × 50 mm2 with a TZM thickness of 5 mm and a graphite thickness of 10 mm have been tested in high heat flux simulation for their thermal fatigue properties. Up to 600 loading cycles were carried out with an average heat flux of 10 MW/m 2 for 0.5 s pulses. The maximum surface temperature was 1100°C. In support of the experiment, the thermal response and temperature gradients of the samples were investigated using a finite element model.

Nickel-Graphite Composite Compliant Interface and/or Hot Shoe Material

NASA Technical Reports Server (NTRS)

Firdosy, Samad A.; Chun-Yip Li, Billy; Ravi, Vilupanur A.; Fleurial, Jean-Pierre; Caillat, Thierry; Anjunyan, Harut

2013-01-01

Next-generation high-temperature thermoelectric-power-generating devices will employ segmented architectures and will have to reliably withstand thermally induced mechanical stresses produced during component fabrication, device assembly, and operation. Thermoelectric materials have typically poor mechanical strength, exhibit brittle behavior, and possess a wide range of coefficient of thermal expansion (CTE) values. As a result, the direct bonding at elevated temperatures of these materials to each other to produce segmented leg components is difficult, and often results in localized microcracking at interfaces and mec hanical failure due to the stresses that arise from the CTE mismatch between the various materials. Even in the absence of full mechanical failure, degraded interfaces can lead to increased electrical and thermal resistances, which adversely impact conversion efficiency and power output. The proposed solution is the insertion of a mechanically compliant layer, with high electrical and thermal conductivity, between the low- and high-temperature segments to relieve thermomechanical stresses during device fabrication and operation. This composite material can be used as a stress-relieving layer between the thermoelectric segments and/or between a thermoelectric segment and a hot- or cold-side interconnect material. The material also can be used as a compliant hot shoe. Nickel-coated graphite powders were hot-pressed to form a nickel-graphite composite material. A freestanding thermoelectric segmented leg was fabricated by brazing the compliant pad layer between the high-temperature p- Zintl and low-temperature p-SKD TE segments using Cu-Ag braze foils. The segmented leg stack was heated in vacuum under a compressive load to achieve bonding. The novelty of the innovation is the use of composite material that re duces the thermomechanical stresses en - countered in the construction of high-efficiency, high-temperature therm - o-electric devices. The compliant pad enables the bonding of dissimilar thermoelectric materials while maintaining the desired electrical and thermal properties essential for efficient device operation. The modulus, CTE, electrical, and thermal conductances of the composite can be controlled by varying the ratio of nickel to graphite.

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

Correlated NanoSIMS, TEM, and XANES Studies of Presolar Grains

NASA Astrophysics Data System (ADS)

Groopman, Evan Edward

The objective of this thesis is to describe the correlated study of individual presolar grains via Nano-scale Secondary Ion Mass Spectrometry (NanoSIMS), Transmission Electron Microscopy (TEM), and Scanning Transmission X-ray Microscopy (STXM) utilizing X-ray Absorption Near Edge Structure (XANES), with a focus on connecting these correlated laboratory studies to astrophysical phenomena. The correlated isotopic, chemical, and microstructural studies of individual presolar grains provide the most detailed description of their formation environments, and help to inform astrophysical models and observations of stellar objects. As a part of this thesis I have developed and improved upon laboratory techniques for micromanipulating presolar grains and embedding them in resin for ultramicrotomy after NanoSIMS analyses and prior to TEM characterization. The new methods have yielded a 100% success rate and allow for the specific correlation of microstructural and isotopic properties of individual grains. Knowing these properties allows for inferences to be made regarding the condensation sequences and the origins of the stellar material that condensed to form these grains. NanoSIMS studies of ultramicrotomed sections of presolar graphite grains have revealed complex isotopic heterogeneities that appear to be primary products of the grains' formation environments and not secondary processing during the grains' lifetimes. Correlated excesses in 15N and 18O were identified as being carried by TiC subgrains within presolar graphite grains from supernovae (SNe). These spatially-correlated isotopic anomalies pinpoint the origin of the material that formed these grains: the inner He/C zone. Complex microstructures and isotopic heterogeneities also provide evidence for mixing in globular SN ejecta, which is corroborated by models and telescopic observations. In addition to these significant isotopic discoveries, I have also observed the first reported nanocrystalline core surrounded by turbostratic graphite within a low-density SN graphite grain. Nanocrystalline cores consisting of randomly-oriented 2-4 nm sheets of graphene and surrounded by concentric shells of graphite have been observed in high-density presolar graphite grains from Asymptotic Giant Branch stars, whose grains are typically microstructurally distinct from SN graphite grains. These vastly different stellar environments briefly formed similar nanocrystalline structures before diverging in the structure of their mantling graphite to be typical of AGB and SN grains. While relatively few correlated NanoSIMS and TEM studies have been performed previously, which this research thesis aims to expand, my collaborators and I also endeavored to add a third correlated technique, STXM/XANES, which had previously not been applied to presolar grains. XANES allows for the investigation of molecular bonds, which we used to help infer physical and chemical properties of stellar ejecta. I investigated the C K-edge and Ti L-edge of molecular bonds in both presolar graphite grains and their TiC subgrains. The presolar graphite grains, while overwhelmingly composed of aromatic C molecules, host a wide variety of minor organic molecules. Considering the large isotopic anomalies in the grains, these minor components are not likely due to contamination. I also investigated the valence state of Ti in Ti-rich subgrains and plan to work towards illuminating the effect that V in solid solution has upon the TiC bonds.

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.

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.

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

10 CFR 110.26 - General license for the export of nuclear reactor components.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear reactor...

10 CFR 110.26 - General license for the export of nuclear reactor components.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear reactor...

10 CFR 110.26 - General license for the export of nuclear reactor components.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear reactor...

10 CFR 110.26 - General license for the export of nuclear reactor components.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear reactor...

10 CFR 110.26 - General license for the export of nuclear reactor components.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear reactor...

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.

Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

NASA Technical Reports Server (NTRS)

Garber, Anne E.; Dickens, Ricky E.

2011-01-01

The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

Supercapacitors based on 3D network of activated carbon nanowhiskers wrapped-on graphitized electrospun nanofibers

NASA Astrophysics Data System (ADS)

He, Shuijian; Chen, Linlin; Xie, Chencheng; Hu, Huan; Chen, Shuiliang; Hanif, Muddasir; Hou, Haoqing

2013-12-01

Due to their cycling stability and high power density, the supercapacitors bridge the power/energy gap between traditional dielectric capacitors and batteries/fuel cells. Electrode materials are key components for making high performance supercapacitors. An activated carbon nanowhiskers (ACNWs) wrapped-on graphitized electrospun nanofiber (GENF) network (ACNWs/GENFN) with 3D porous structure is prepared as a new type of binder-free electrode material for supercapacitors. The supercapacitor based on the ACNWs/GENFN composite material displays an excellent performance with a specific capacitance of 176.5 F g-1 at current density of 0.5 A g-1, an ultrahigh power density of 252.8 kW kg-1 at current density of 800 A g-1 and an outstanding cycling stability of no capacitance loss after 10,000 charge/discharge cycles.

Capacitance‐Assisted Sustainable Electrochemical Carbon Dioxide Mineralisation

PubMed Central

Lamb, Katie J.; Dowsett, Mark R.; Chatzipanagis, Konstantinos; Scullion, Zhan Wei; Kröger, Roland; Lee, James D.

2017-01-01

Abstract An electrochemical cell comprising a novel dual‐component graphite and Earth‐crust abundant metal anode, a hydrogen producing cathode and an aqueous sodium chloride electrolyte was constructed and used for carbon dioxide mineralisation. Under an atmosphere of 5 % carbon dioxide in nitrogen, the cell exhibited both capacitive and oxidative electrochemistry at the anode. The graphite acted as a supercapacitive reagent concentrator, pumping carbon dioxide into aqueous solution as hydrogen carbonate. Simultaneous oxidation of the anodic metal generated cations, which reacted with the hydrogen carbonate to give mineralised carbon dioxide. Whilst conventional electrochemical carbon dioxide reduction requires hydrogen, this cell generates hydrogen at the cathode. Carbon capture can be achieved in a highly sustainable manner using scrap metal within the anode, seawater as the electrolyte, an industrially relevant gas stream and a solar panel as an effective zero‐carbon energy source. PMID:29171724

α-MnO2 nanorods supported on porous graphitic carbon nitride as efficient electrocatalysts for lithium-air batteries

NASA Astrophysics Data System (ADS)

Hang, Yang; Zhang, Chaofeng; Luo, Xiaoman; Xie, Yingshen; Xin, Sen; Li, Yutao; Zhang, Dawei; Goodenough, John B.

2018-07-01

Synthesis of α-MnO2 nanorods grown on porous graphitic carbon nitride (g-C3N4) sheets via a facile hydrothermal treatment gives a porous composite exhibiting higher activity for an air cathode than the individual component of α-MnO2 or porous g-C3N4 for both the oxygen reduction reaction (ORR) and oxygen evolution reaction (OER). The porous g-C3N4/α-MnO2 composite also exhibits better performance in a Li-air battery than pure α-MnO2 or XC-72 carbon catalysts, which includes superior discharge capacity, low voltage gap and high cycle stability. The α-MnO2 nanorods catalyze the OER and the porous g-C3N4 sheets catalyze the ORR.

Graphene in NLO Devices for High Energy Laser Protection

DTIC Science & Technology

2009-11-17

for industrial applications, has been working to advance the application base of graphene . We have recently demonstrated in laser protection...component for evaluation and use of graphene suspensions for laser protection is dispersion of the graphene sheets into appropriate solvents... graphene sheets peeled off from graphite with scotch-tape. For applications where industrial quantities of graphene are needed, however

Engineering report. Part 3: NASA lightweight wheel and brake sub-system. Lightweight brake development. [for application to space shuttle

NASA Technical Reports Server (NTRS)

Bok, L. D.

1973-01-01

The development of light weight wheel and brake systems designed to meet the space shuttle type requirements was investigated. The study includes the use of carbon graphite composite and beryllium as heat sink materials and the compatibility of these heat sink materials with the other structural components of the wheel and brake.

KENNEDY SPACE CENTER, FLA. - In the Orbiter Processing Facility, an orbital maneuvering system (OMS) pod is moved into place on Atlantis. It is one of two OMS pods attached to the upper aft fuselage left and right sides. Fabricated primarily of graphite epoxy composite and aluminum, each pod is 21.8 feet long and 11.37 feet wide at its aft end and 8.41 feet wide at its forward end, with a surface area of approximately 435 square feet. Each pod houses the Reaction Control System propulsion components used for inflight maneuvering and is attached to the aft fuselage with 11 bolts.

NASA Image and Video Library

2003-10-30

KENNEDY SPACE CENTER, FLA. - In the Orbiter Processing Facility, an orbital maneuvering system (OMS) pod is moved into place on Atlantis. It is one of two OMS pods attached to the upper aft fuselage left and right sides. Fabricated primarily of graphite epoxy composite and aluminum, each pod is 21.8 feet long and 11.37 feet wide at its aft end and 8.41 feet wide at its forward end, with a surface area of approximately 435 square feet. Each pod houses the Reaction Control System propulsion components used for inflight maneuvering and is attached to the aft fuselage with 11 bolts.

KENNEDY SPACE CENTER, FLA. - In the Orbiter Processing Facility, an orbital maneuvering system (OMS) pod is suspended in air as it is moved toward Atlantis for installation. Two OMS pods are attached to the upper aft fuselage left and right sides. Fabricated primarily of graphite epoxy composite and aluminum, each pod is 21.8 feet long and 11.37 feet wide at its aft end and 8.41 feet wide at its forward end, with a surface area of approximately 435 square feet. Each pod houses the Reaction Control System propulsion components used for inflight maneuvering and is attached to the aft fuselage with 11 bolts.

NASA Image and Video Library

2003-10-30

KENNEDY SPACE CENTER, FLA. - In the Orbiter Processing Facility, an orbital maneuvering system (OMS) pod is suspended in air as it is moved toward Atlantis for installation. Two OMS pods are attached to the upper aft fuselage left and right sides. Fabricated primarily of graphite epoxy composite and aluminum, each pod is 21.8 feet long and 11.37 feet wide at its aft end and 8.41 feet wide at its forward end, with a surface area of approximately 435 square feet. Each pod houses the Reaction Control System propulsion components used for inflight maneuvering and is attached to the aft fuselage with 11 bolts.

KENNEDY SPACE CENTER, FLA. - In the Orbiter Processing Facility, an orbital maneuvering system (OMS) pod is moved closer to Atlantis for installation. Two OMS pods are attached to the upper aft fuselage left and right sides. Fabricated primarily of graphite epoxy composite and aluminum, each pod is 21.8 feet long and 11.37 feet wide at its aft end and 8.41 feet wide at its forward end, with a surface area of approximately 435 square feet. Each pod houses the Reaction Control System propulsion components used for inflight maneuvering and is attached to the aft fuselage with 11 bolts.

NASA Image and Video Library

2003-10-30

KENNEDY SPACE CENTER, FLA. - In the Orbiter Processing Facility, an orbital maneuvering system (OMS) pod is moved closer to Atlantis for installation. Two OMS pods are attached to the upper aft fuselage left and right sides. Fabricated primarily of graphite epoxy composite and aluminum, each pod is 21.8 feet long and 11.37 feet wide at its aft end and 8.41 feet wide at its forward end, with a surface area of approximately 435 square feet. Each pod houses the Reaction Control System propulsion components used for inflight maneuvering and is attached to the aft fuselage with 11 bolts.

KENNEDY SPACE CENTER, FLA. - In the Orbiter Processing Facility, technicians make adjustments to the orbital maneuvering system (OMS) pod being installed on Atlantis. The OMS pod is one of two that are attached to the upper aft fuselage left and right sides. Fabricated primarily of graphite epoxy composite and aluminum, each pod is 21.8 feet long and 11.37 feet wide at its aft end and 8.41 feet wide at its forward end, with a surface area of approximately 435 square feet. Each pod houses the Reaction Control System propulsion components used for inflight maneuvering and is attached to the aft fuselage with 11 bolts.

NASA Image and Video Library

2003-10-30

KENNEDY SPACE CENTER, FLA. - In the Orbiter Processing Facility, technicians make adjustments to the orbital maneuvering system (OMS) pod being installed on Atlantis. The OMS pod is one of two that are attached to the upper aft fuselage left and right sides. Fabricated primarily of graphite epoxy composite and aluminum, each pod is 21.8 feet long and 11.37 feet wide at its aft end and 8.41 feet wide at its forward end, with a surface area of approximately 435 square feet. Each pod houses the Reaction Control System propulsion components used for inflight maneuvering and is attached to the aft fuselage with 11 bolts.

KENNEDY SPACE CENTER, FLA. - In the Orbiter Processing Facility, one of two orbital maneuvering system (OMS) pods is being moved for installation on Atlantis. The OMS pods are attached to the upper aft fuselage left and right sides. Fabricated primarily of graphite epoxy composite and aluminum, each pod is 21.8 feet long and 11.37 feet wide at its aft end and 8.41 feet wide at its forward end, with a surface area of approximately 435 square feet. Each pod houses the Reaction Control System propulsion components used for inflight maneuvering and is attached to the aft fuselage with 11 bolts.

NASA Image and Video Library

2003-10-30

KENNEDY SPACE CENTER, FLA. - In the Orbiter Processing Facility, one of two orbital maneuvering system (OMS) pods is being moved for installation on Atlantis. The OMS pods are attached to the upper aft fuselage left and right sides. Fabricated primarily of graphite epoxy composite and aluminum, each pod is 21.8 feet long and 11.37 feet wide at its aft end and 8.41 feet wide at its forward end, with a surface area of approximately 435 square feet. Each pod houses the Reaction Control System propulsion components used for inflight maneuvering and is attached to the aft fuselage with 11 bolts.

KENNEDY SPACE CENTER, FLA. - In the Orbiter Processing Facility, technicians move an orbital maneuvering system (OMS) pod into the correct position on Atlantis. The OMS pod is one of two that are attached to the upper aft fuselage left and right sides. Fabricated primarily of graphite epoxy composite and aluminum, each pod is 21.8 feet long and 11.37 feet wide at its aft end and 8.41 feet wide at its forward end, with a surface area of approximately 435 square feet. Each pod houses the Reaction Control System propulsion components used for inflight maneuvering and is attached to the aft fuselage with 11 bolts.

NASA Image and Video Library

2003-10-30

KENNEDY SPACE CENTER, FLA. - In the Orbiter Processing Facility, technicians move an orbital maneuvering system (OMS) pod into the correct position on Atlantis. The OMS pod is one of two that are attached to the upper aft fuselage left and right sides. Fabricated primarily of graphite epoxy composite and aluminum, each pod is 21.8 feet long and 11.37 feet wide at its aft end and 8.41 feet wide at its forward end, with a surface area of approximately 435 square feet. Each pod houses the Reaction Control System propulsion components used for inflight maneuvering and is attached to the aft fuselage with 11 bolts.

Effects of Nuclear Weapons.

ERIC Educational Resources Information Center

Sartori, Leo

1983-01-01

Fundamental principles governing nuclear explosions and their effects are discussed, including three components of a nuclear explosion (thermal radiation, shock wave, nuclear radiation). Describes how effects of these components depend on the weapon's yield, its height of burst, and distance of detonation point. Includes effects of three…

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.

Graphite fiber textile preform/copper matrix composites

NASA Technical Reports Server (NTRS)

Gilatovs, G. J.; Lee, Bruce; Bass, Lowell

1995-01-01

Graphite fiber reinforced/copper matrix composites have sufficiently high thermal conduction to make them candidate materials for critical heat transmitting and rejection components. The term textile composites arises because the preform is braided from fiber tows, conferring three-dimensional reinforcement and near net shape. The principal issues investigated in the past two years have centered on developing methods to characterize the preform and fabricated composite and on braidability. It is necessary to have an analytic structural description for both processing and final property modeling. The structure of the true 3-D braids used is complex and has required considerable effort to model. A structural mapping has been developed as a foundation for analytic models for thermal conduction and mechanical properties. The conductivity has contributions both from the copper and the reinforcement. The latter is accomplished by graphitization of the fibers, the higher the amount of graphitization the greater the conduction. This is accompanied by an increase in the fiber modulus, which is desirable from a stiffness point of view but decreases the braidability; the highest conductivity fibers are simply too brittle to be braided. Considerable effort has been expended on determining the optimal braidability--conductivity region. While a number of preforms have been fabricated, one other complication intervenes; graphite and copper are immiscible, resulting in a poor mechanical bond and difficulties in infiltration by molten copper. The approach taken is to utilize a proprietary fiber coating process developed by TRA, of Salt Lake City, Utah, which forms an itermediary bond. A number of preforms have been fabricated from a variety of fiber types and two sets of these have been infiltrated with OFHC copper, one with the TRA coating and one without. Mechanical tests have been performed using a small-scale specimen method and show the coated specimens to have superior mechanical properties. Final batches of preforms, including a finned, near net shape tube, are being fabricated and will be infiltrated before summer.

Virtual enterprise model for the electronic components business in the Nuclear Weapons Complex

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

Ferguson, T.J.; Long, K.S.; Sayre, J.A.

1994-08-01

The electronic components business within the Nuclear Weapons Complex spans organizational and Department of Energy contractor boundaries. An assessment of the current processes indicates a need for fundamentally changing the way electronic components are developed, procured, and manufactured. A model is provided based on a virtual enterprise that recognizes distinctive competencies within the Nuclear Weapons Complex and at the vendors. The model incorporates changes that reduce component delivery cycle time and improve cost effectiveness while delivering components of the appropriate quality.

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.

The state of understanding of the lithium-ion-battery graphite solid electrolyte interphase (SEI) and its relationship to formation cycling

DOE PAGES

An, Seong Jin; Li, Jianlin; Daniel, Claus; ...

2016-04-09

An in-depth review is presented on the science of lithium-ion battery (LIB) solid electrolyte interphase (SEI) formation on the graphite anode, including structure, morphology, chemical composition, electrochemistry, formation mechanism, and LIB formation cycling. During initial operation of LIBs, the SEI layer forms on the graphite surfaces, the most commonly used anode material, due to side reactions with the electrolyte solvent/salt at low electro-reduction potentials. It is accepted that the SEI layer is essential to the long-term performance of LIBs, and it also has an impact on its initial capacity loss, self-discharge characteristics, cycle life, rate capability, and safety. While themore » presence of the anode SEI layer is vital, it is difficult to control its formation and growth, as the chemical composition, morphology, and stability depend on several factors. These factors include the type of graphite, electrolyte composition, electrochemical conditions, and cell temperature. Thus, SEI layer formation and electrochemical stability over long-term operation should be a primary topic of future investigation in the development of LIB technology. We review the progression of knowledge gained about the anode SEI, from its discovery in 1979 to the current state of understanding, and covers its formation process, differences in the chemical and structural makeup when cell materials and components are varied, methods of characterization, and associated reactions with the liquid electrolyte phase. It also discusses the relationship of the SEI layer to the LIB formation step, which involves both electrolyte wetting and subsequent slow charge-discharge cycles to grow the SEI.« less

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.

Improving nuclear envelope dynamics by EBV BFRF1 facilitates intranuclear component clearance through autophagy.

PubMed

Liu, Guan-Ting; Kung, Hsiu-Ni; Chen, Chung-Kuan; Huang, Cheng; Wang, Yung-Li; Yu, Cheng-Pu; Lee, Chung-Pei

2018-02-26

Although a vesicular nucleocytoplasmic transport system is believed to exist in eukaryotic cells, the features of this pathway are mostly unknown. Here, we report that the BFRF1 protein of the Epstein-Barr virus improves vesicular transport of nuclear envelope (NE) to facilitate the translocation and clearance of nuclear components. BFRF1 expression induces vesicles that selectively transport nuclear components to the cytoplasm. With the use of aggregation-prone proteins as tools, we found that aggregated nuclear proteins are dispersed when these BFRF1-induced vesicles are formed. BFRF1-containing vesicles engulf the NE-associated aggregates, exit through from the NE, and putatively fuse with autophagic vacuoles. Chemical treatment and genetic ablation of autophagy-related factors indicate that autophagosome formation and autophagy-linked FYVE protein-mediated autophagic proteolysis are involved in this selective clearance of nuclear proteins. Remarkably, vesicular transport, elicited by BFRF1, also attenuated nuclear aggregates accumulated in neuroblastoma cells. Accordingly, induction of NE-derived vesicles by BFRF1 facilitates nuclear protein translocation and clearance, suggesting that autophagy-coupled transport of nucleus-derived vesicles can be elicited for nuclear component catabolism in mammalian cells.-Liu, G.-T., Kung, H.-N., Chen, C.-K., Huang, C., Wang, Y.-L., Yu, C.-P., Lee, C.-P. Improving nuclear envelope dynamics by EBV BFRF1 facilitates intranuclear component clearance through autophagy.

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.

The nature and origin of interstellar diamond

NASA Technical Reports Server (NTRS)

Blake, David F.; Freund, Friedemann; Shipp, Ruth; Krishnan, Kannan F. M.; Echer, Charles J.

1988-01-01

The C-delta component of the Allende meteorite is a microscopic diamond some of whose properties seem in conflict with those expected of diamond. High spatial resolution analytical data are presented here which may help explain such results. Surface and interfacial carbon atoms in the component, which may comprise as much as 25 percent of the total, impart an 'amorphous' character to some spectral data. These data support the proposed high-pressure conversion of amorphous carbon and graphite into diamonds due to grain-grain collisions in the ISM, although a low-pressure mechanism of formation cannot be ruled out.

Metallic Wall Hall Thrusters

NASA Technical Reports Server (NTRS)

Goebel, Dan Michael (Inventor); Hofer, Richard Robert (Inventor); Mikellides, Ioannis G. (Inventor)

2016-01-01

A Hall thruster apparatus having walls constructed from a conductive material, such as graphite, and having magnetic shielding of the walls from the ionized plasma has been demonstrated to operate with nearly the same efficiency as a conventional non-magnetically shielded design using insulators as wall components. The new design is believed to provide the potential of higher power and uniform operation over the operating life of a thruster device.

Metallic Wall Hall Thrusters

NASA Technical Reports Server (NTRS)

Goebel, Dan Michael (Inventor); Hofer, Richard Robert (Inventor); Mikellides, Ioannis G. (Inventor)

2018-01-01

A Hall thruster apparatus having walls constructed from a conductive material, such as graphite, and having magnetic shielding of the walls from the ionized plasma has been demonstrated to operate with nearly the same efficiency as a conventional nonmagnetically shielded design using insulators as wall components. The new design is believed to provide the potential of higher power and uniform operation over the operating life of a thruster device.

Design and construction of a Faraday cup for measurement of small electronic currents

NASA Technical Reports Server (NTRS)

Veyssiere, A.

1985-01-01

The design of a device to measure and integrate very small currents generated by the impact of a charged particle beam upon a Faraday cut is described. The main component is a graphite block capable of stopping practically all the incident changes. The associated electronic apparatus required to measure better than 10/13 ampere with a precision of 10/0 is described.

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

Ammigan, K.; Hurh, P.

The Radiation Damage In Accelerator Target Environments (RaDIATE) collaboration was founded in 2012 and currently consists of over 50 participants and 11 institutions globally. Due to the increasing power of future proton accelerator sources in target facilities, there is a critical need to further understand the physical and thermo-mechanical radiation response of target facility materials. Thus, the primary objective of the RaDIATE collaboration is to draw on existing expertise in the nuclear materials and accelerator targets fields to generate new and useful materials data for application within the accelerator and fission/fusion communities. Current research activities of the collaboration include postmore » irradiation examination (PIE) of decommissioned components from existing beamlines such as the NuMI beryllium beam window and graphite NT-02 target material. PIE of these components includes advanced microstructural analyses (SEM/TEM, EBSD, EDS) and micro-mechanics technique such as nano-indentation, to help characterize any microstructural radiation damage incurred during operation. New irradiation campaigns of various candidate materials at both low and high energy beam facilities are also being pursued. Beryllium helium implantation studies at the University of Surrey as well as high energy proton irradiation of various materials at Brookhaven National Laboratory’s BLIP facility have been initiated. The program also extends to beam-induced thermal shock experiments using high intensity beam pulses at CERN’s HiRadMat facility, followed by advanced PIE activities to evaluate thermal shock resistance of the materials. Preliminary results from ongoing research activities, as well as the future plans of the RaDIATE collaboration R&D program will be discussed.« less

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.

Constraints on Grain Formation Around Carbon Stars from Laboratory Studies of Presolar Graphite

NASA Technical Reports Server (NTRS)

Bernatowicz, T. J.; Akande, O. W.; Croat, T. K.; Cowsik, R.

2005-01-01

We report the results of an investigation into the physical conditions in the mass outflows of asymptotic giant branch (AGB) carbon stars that are required for the formation of micron-sized presolar graphite grains, either with or without internal crystals of titanium carbide (TiC). In addition to providing detailed information about stellar nucleosynthesis, the structure and composition of presolar grains give unique information about the conditions of grain formation. In the present work we use laboratory observations of presolar graphite to gain insight into the physical conditions in circumstellar outflows from carbon AGB stars. The periodic pulsation of AGB stars enhances the gas density through shocks in the stellar atmosphere above the photosphere, promoting the condensation of dust grains. Copious mass outflow occurs largely because grains are coupled to the radiation field of the star, which accelerates them by radiation pressure; momentum is in turn transferred to gas molecules by collisions with grains. The dust/gas mixture is effectively a two-component fluid whose motion depends on atmospheric structure and which, in turn, influences that structure. In particular, the radiation pressure on the grains determines the velocity field of the outflow and thus the density distribution, while the density distribution itself determines the conditions of radiative transfer within the outflow and thus the effective radiation pressure.

Breakdown voltage reliability improvement in gas-discharge tube surge protectors employing graphite field emitters

NASA Astrophysics Data System (ADS)

Žumer, Marko; Zajec, Bojan; Rozman, Robert; Nemanič, Vincenc

2012-04-01

Gas-discharge tube (GDT) surge protectors are known for many decades as passive units used in low-voltage telecom networks for protection of electrical components from transient over-voltages (discharging) such as lightning. Unreliability of the mean turn-on DC breakdown voltage and the run-to-run variability has been overcome successfully in the past by adding, for example, a radioactive source inside the tube. Radioisotopes provide a constant low level of free electrons, which trigger the breakdown. In the last decades, any concept using environmentally harmful compounds is not acceptable anymore and new solutions were searched. In our application, a cold field electron emitter source is used as the trigger for the gas discharge but with no activating compound on the two main electrodes. The patent literature describes in details the implementation of the so-called trigger wires (auxiliary electrodes) made of graphite, placed in between the two main electrodes, but no physical explanation has been given yet. We present experimental results, which show that stable cold field electron emission current in the high vacuum range originating from the nano-structured edge of the graphite layer is well correlated to the stable breakdown voltage of the GDT surge protector filled with a mixture of clean gases.

Development of fuel cell bipolar plates from graphite filled wet-lay thermoplastic composite materials

NASA Astrophysics Data System (ADS)

Huang, Jianhua; Baird, Donald G.; McGrath, James E.

A method with the potential to produce economical bipolar plates with high electrical conductivity and mechanical properties is described. Thermoplastic composite materials consisting of graphite particles, thermoplastic fibers and glass or carbon fibers are generated by means of a wet-lay (paper-making) process to yield highly formable sheets. The sheets are then stacked and compression molded to form bipolar plates with gas flow channels. Poly(phenylene sulfide) (PPS) based wet-lay composite plates have in-plane conductivity of 200-300 S cm -1, tensile strength of 57 MPa, flexural strength of 96 MPa and impact strength (unnotched) of 81 J m -1 (1.5 ft-lb in. -1). These values well exceed industrial as well as Department of Energy requirements or targets and have never been reached before for composite bipolar plates. The use of wet-lay sheets also makes it possible to choose different components including polymer, graphite particle and reinforcement for the core and outer layers of the plate, respectively, to optimize the properties and/or reduce the cost of the plate. The through-plane conductivity (around 20 S cm -1) and half-cell resistance of the bipolar plate indicate that the through-plane conductivity of the material needs some improvement.

Initial Gamma Spectrometry Examination of the AGR-3/4 Irradiation

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

Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

2016-11-01

The initial results from gamma spectrometry examination of the different components from the combined third and fourth US Advanced Gas Reactor Fuel Development TRISO-coated particle fuel irradiation tests (AGR-3/4) have been analyzed. This experiment was designed to provide information about in-pile fission product migration. In each of the 12 capsules, a single stack of four compacts with designed-to-fail particles surrounded by two graphitic diffusion rings (inner and outer) and a graphite sink were irradiated in the Idaho National Laboratory’s Advanced Test Reactor. Gamma spectrometry has been used to evaluate the gamma-emitting fission product inventory of compacts from the irradiation andmore » evaluate the burnup of these compacts based on the activity of the radioactive cesium isotopes (Cs-134 and Cs-137) in the compacts. Burnup from gamma spectrometry compares well with predicted burnup from simulations. Additionally, inner and outer rings were also examined by gamma spectrometry both to evaluate the fission product inventory and the distribution of gamma-emitting fission products within the rings using gamma emission computed tomography. The cesium inventory of the scanned rings compares acceptably well with the expected inventory from fission product transport modeling. The inventory of the graphite fission product sinks is also being evaluated by gamma spectrometry.« less

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.

Synthesis of Multimetal-Graphene Composite by Mechanical Milling

NASA Astrophysics Data System (ADS)

Saiphaneendra, Bachu; Srivastava, Avi Krishna; Srivastava, Chandan

2016-10-01

Multimetal-graphene composites were synthesized using the ball milling technique. To prepare the composite, graphite powder was mixed with Fe, Cr, Co, Cu and Mg powders. This mixture was then mechanically milled for 35 h in toluene medium. After milling, the multimetal-graphite mixture was mixed with sodium lauryl sulfate and sonicated for 2 h. Sonication led to the exfoliation of graphene sheets. Formation of graphene was confirmed from x-ray diffraction and Raman spectroscopy. Transmission electron microscopy-based analysis revealed the formation of multimetal deposits over the graphene surface. Compositional analysis of the multimetal deposits revealed fairly uniform distribution of all the five component metal atoms over the graphene sheet. The average composition of the multimetal deposit was determined to be 11.4 ± 4 at.% Mg, 33.8 ± 19 at.% Cr, 21.8 ± 16 at.% Fe, 9.4 ± 5.7 at.% Co and 23.6 ± 12 at.% Cu.

Arcing and its role in PFC erosion and dust production in DIII-D

NASA Astrophysics Data System (ADS)

Rudakov, D. L.; Chrobak, C. P.; Doerner, R. P.; Krasheninnikov, S. I.; Moyer, R. A.; Umstadter, K. R.; Wampler, W. R.; Wong, C. P. C.

2013-07-01

Two types of arc tracks are observed on the plasma-facing components (PFCs) in DIII-D. "Unmagnetized" random walk tracks are produced during glow discharges; they are rare and have no importance for PFC erosion but may degrade diagnostic mirrors. "Magnetized" scratch-like type II tracks are produced by unipolar arcs during plasma operations; they are formed by "retrograde BxJ" motion of the cathode spot and are roughly perpendicular to the local magnetic field. Type II arcs cause measurable erosion of graphite, but based on the evidence available they are relatively small contributors to the total erosion of carbon in DIII-D compared to other mechanisms such as physical and chemical sputtering and ablation from leading edges. Erosion by arcing of tungsten films deposited on graphite samples was observed in Divertor Material Evaluation System (DiMES) experiments. New DiMES experiments aimed at time-resolved arc measurements are proposed.

Nafion induced surface confinement of oxygen in carbon-supported oxygen reduction catalysts

DOE PAGES

Chlistunoff, Jerzy; Sansinena, Jose -Maria

2016-11-17

We studied the surface confinement of oxygen inside layers of Nafion self-assembled on carbon-supported oxygen reduction reaction (ORR) catalysts. It is demonstrated that oxygen accumulates in the hydrophobic component of the polymer remaining in contact with the carbon surface. Furthermore, the amount of surface confined oxygen increases with the degree of carbon surface graphitization, which promotes the self-assembly of the polymer. Planar macrocyclic ORR catalysts possessing a delocalized system of π electrons such as Co and Fe porphyrins and phthalocyanines have virtually no effect on the surface confinement of oxygen, in accordance with their structural similarity to graphitic carbon surfacesmore » where they adsorb. Platinum particles in carbon-supported ORR catalysts with high metal contents (20%) disrupt the self-assembly of Nafion and virtually eliminate the oxygen confinement, but the phenomenon is still observed for low Pt loading (4.8%) catalysts.« less

Study to investigate design, fabrication and test of low cost concepts for large hybrid composite helicopter fuselage, phase 2

NASA Technical Reports Server (NTRS)

Adams, K. M.; Lucas, J. J.

1977-01-01

The development of a frame/stringer/skin fabrication technique for composite airframe construction was studied as a low cost approach to the manufacturer of larger helicopter airframe components. A center cabin aluminum airframe section of the Sikorsky CH-53D, was selected for evaluation as a composite structure. The design, as developed, is composed of a woven KEVLAR R-49/epoxy skin and graphite/epoxy frames and stringers. The single cure concept is made possible by the utilization of pre-molded foam cores, over which the graphite/epoxy pre-impregnated frame and stringer reinforcements are positioned. Bolted composite channel sections were selected as the optimum joint construction. The applicability of the single cure concept to larger realistic curved airframe sections, and the durability of the composite structure in a realistic spectrum fatigue environment, was described.

Repair techniques for celion/LARC-160 graphite/polyimide composite structures

NASA Technical Reports Server (NTRS)

Jones, J. S.; Graves, S. R.

1984-01-01

The large stiffness-to-weight and strength-to-weight ratios of graphite composite in combination with the 600 F structural capability of the polyimide matrix can reduce the total structure/TPS weight of reusable space vehicles by 20-30 percent. It is inevitable that with planned usage of GR/PI structural components, damage will occur either in the form of intrinsic flaw growth or mechanical damage. Research and development programs were initiated to develop repair processes and techniques specific to Celion/LARC-160 GR/PI structure with emphasis on highly loaded and lightly loaded compression critical structures for factory type repair. Repair processes include cocure and secondary bonding techniques applied under vacuum plus positive autoclave pressure. Viable repair designs and processes are discussed for flat laminates, honeycomb sandwich panels, and hat-stiffened skin-stringer panels. The repair methodology was verified through structural element compression tests at room temperature and 315 C (600 F).

Surface Structure of Aerobically Oxidized Diamond Nanocrystals

DOE PAGES

Wolcott, Abraham; Schiros, Theanne; Trusheim, Matthew E.; ...

2014-10-27

Here we investigate the aerobic oxidation of high-pressure, high-temperature nanodiamonds (5–50 nm dimensions) using a combination of carbon and oxygen K-edge X-ray absorption, wavelength-dependent X-ray photoelectron, and vibrational spectroscopies. Oxidation at 575 °C for 2 h eliminates graphitic carbon contamination (>98%) and produces nanocrystals with hydroxyl functionalized surfaces as well as a minor component (<5%) of carboxylic anhydrides. The low graphitic carbon content and the high crystallinity of HPHT are evident from Raman spectra acquired using visible wavelength excitation (λ excit = 633 nm) as well as carbon K-edge X-ray absorption spectra where the signature of a core–hole exciton ismore » observed. Both spectroscopic features are similar to those of chemical vapor deposited (CVD) diamond but differ significantly from the spectra of detonation nanodiamond. Lastly, we discuss the importance of these findings to the functionalization of nanodiamond surfaces for biological labeling applications.« less

Molten Salt Electrolytically Produced Carbon/Tin Nanomaterial as the Anode in a Lithium Ion Battery

NASA Astrophysics Data System (ADS)

Das Gupta, Rajshekar; Schwandt, Carsten; Fray, Derek J.

2017-03-01

A carbon/tin nanomaterial, consisting of predominantly Sn-filled carbon nanotubes and nanoparticles, is prepared by molten salt electrochemistry, using electrodes of graphite and an electrolyte of LiCl salt containing a small admixture of SnCl2. The C/Sn hybrid material generated is incorporated into the active anode material of a lithium ion battery and tested with regard to storage capacity and cycling behavior. The results demonstrate that the C/Sn material has favorable properties, in terms of energy density and in particular long-term stability, that exceed those of the individual components alone. The initial irreversible capacity of the material is somewhat larger than that of conventional battery graphite which is due to its unique nanostructure. Overall the results would indicate the suitability of this material for use in the anodes of lithium ion batteries with high rate capability.

Design, fabrication and test of graphite/polyimide composite joints and attachments for advanced aerospace vehicles

NASA Technical Reports Server (NTRS)

1981-01-01

The development of several types of graphite/polyimide (GR/PI) bonded and bolted joints is reported. The program consists of two concurrent tasks: (1) design and test of specific built up attachments; and (2) evaluation of standard advanced bonded joint concepts. A data base for the design and analysis of advanced composite joints for use at elevated temperatures (561K (550 deg F)) to design concepts for specific joining applications, and the fundamental parameters controlling the static strength characteristics of such joints are evaluated. Data for design and build GR/PI of lightly loaded flight components for advanced space transportation systems and high speed aircraft are presented. Results for compression and interlaminar shear strengths of Celion 6000/PMR-15 laminates are given. Static discriminator test results for type 3 and type 4 bonded and bolted joints and final joint designs for TASK 1.4 scale up fabrication and testing are presented.

Surface Structure of Aerobically Oxidized Diamond Nanocrystals

PubMed Central

2015-01-01

We investigate the aerobic oxidation of high-pressure, high-temperature nanodiamonds (5–50 nm dimensions) using a combination of carbon and oxygen K-edge X-ray absorption, wavelength-dependent X-ray photoelectron, and vibrational spectroscopies. Oxidation at 575 °C for 2 h eliminates graphitic carbon contamination (>98%) and produces nanocrystals with hydroxyl functionalized surfaces as well as a minor component (<5%) of carboxylic anhydrides. The low graphitic carbon content and the high crystallinity of HPHT are evident from Raman spectra acquired using visible wavelength excitation (λexcit = 633 nm) as well as carbon K-edge X-ray absorption spectra where the signature of a core–hole exciton is observed. Both spectroscopic features are similar to those of chemical vapor deposited (CVD) diamond but differ significantly from the spectra of detonation nanodiamond. The importance of these findings to the functionalization of nanodiamond surfaces for biological labeling applications is discussed. PMID:25436035

Capacitance-Assisted Sustainable Electrochemical Carbon Dioxide Mineralisation.

PubMed

Lamb, Katie J; Dowsett, Mark R; Chatzipanagis, Konstantinos; Scullion, Zhan Wei; Kröger, Roland; Lee, James D; Aguiar, Pedro M; North, Michael; Parkin, Alison

2018-01-10

An electrochemical cell comprising a novel dual-component graphite and Earth-crust abundant metal anode, a hydrogen producing cathode and an aqueous sodium chloride electrolyte was constructed and used for carbon dioxide mineralisation. Under an atmosphere of 5 % carbon dioxide in nitrogen, the cell exhibited both capacitive and oxidative electrochemistry at the anode. The graphite acted as a supercapacitive reagent concentrator, pumping carbon dioxide into aqueous solution as hydrogen carbonate. Simultaneous oxidation of the anodic metal generated cations, which reacted with the hydrogen carbonate to give mineralised carbon dioxide. Whilst conventional electrochemical carbon dioxide reduction requires hydrogen, this cell generates hydrogen at the cathode. Carbon capture can be achieved in a highly sustainable manner using scrap metal within the anode, seawater as the electrolyte, an industrially relevant gas stream and a solar panel as an effective zero-carbon energy source. © 2017 The Authors. Published by Wiley-VCH Verlag GmbH & Co. KGaA.

Nafion induced surface confinement of oxygen in carbon-supported oxygen reduction catalysts

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

Chlistunoff, Jerzy; Sansinena, Jose -Maria

We studied the surface confinement of oxygen inside layers of Nafion self-assembled on carbon-supported oxygen reduction reaction (ORR) catalysts. It is demonstrated that oxygen accumulates in the hydrophobic component of the polymer remaining in contact with the carbon surface. Furthermore, the amount of surface confined oxygen increases with the degree of carbon surface graphitization, which promotes the self-assembly of the polymer. Planar macrocyclic ORR catalysts possessing a delocalized system of π electrons such as Co and Fe porphyrins and phthalocyanines have virtually no effect on the surface confinement of oxygen, in accordance with their structural similarity to graphitic carbon surfacesmore » where they adsorb. Platinum particles in carbon-supported ORR catalysts with high metal contents (20%) disrupt the self-assembly of Nafion and virtually eliminate the oxygen confinement, but the phenomenon is still observed for low Pt loading (4.8%) catalysts.« less

Research Technology

NASA Image and Video Library

1998-01-01

Engineers at Marshall Space Flight Center (MSFC) in Huntsville, Alabama, are working with industry partners to develop a new generation of more cost-efficient space vehicles. Lightweight fuel tanks and components under development will be the critical elements in tomorrow's reusable launch vehicles and will tremendously curb the costs of getting to space. In this photo, Tom DeLay, a materials processes engineer for MSFC, uses a new graphite epoxy technology to create lightweight cryogenic fuel lines for futuristic reusable launch vehicles. He is wrapping a water-soluble mandrel, or mold, with a graphite fabric coated with an epoxy resin. Once wrapped, the pipe will be vacuum-bagged and autoclave-cured. The disposable mold will be removed to reveal a thin-walled fuel line. In addition to being much lighter and stronger than metal, this material won't expand or contract as much in the extreme temperatures encountered by launch vehicles.

Development and demonstration of manufacturing processes for fabricating graphite/Larc-160 polyimide structural elements, part 4, paragraph C

NASA Technical Reports Server (NTRS)

1981-01-01

Progress in the development of processes for production of Celion/LARC-160 graphite-polyimide materials, quality control methods, and the fabrication of Space Shuttle composite structure components is reported. The formulation and processing limits for three batches of resin are presented. Process improvements for simplification of the imidizing and autoclave cure cycles are described. Imidized and autoclave cured test panels were prepared. Celion/LARC-160 cure process verification and the fabrication of honeycomb sandwich panel elements and skin/stringer panels are described. C-scans of laminates imidized at 163 C to 218 C for periods from 30 to 180 minutes, and of process verification laminates made from different batches of prepreg are presented. Failure modes and load/strain characteristics of sandwich elements and C-scans of stringer to skin bond joints are also given.

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