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At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Sliding friction and wear behavior of nuclear graphite in high temperature inert environment: Influence of contact load, speed and temperature

Repeated dynamic interactions of graphitic components in pebble-bed gas-cooled nuclear reactors can cause abrasive wear-induced pebble surface damage, generate hazardous fine graphite debris, and alter fuel circulation dynamics due to changes in friction behavior. Comprehensive tribological characterization of nuclear graphitic materials in conditions relevant to reactor operation is needed to assess reactor long-term safety and performance. This work reports sliding friction and wear behavior of self-mated nuclear graphite ET-10 at various elevated temperatures (650 °C and 750 °C), sliding speeds (1 and 10 mm/s) and contact loads (20 and 40 N) in a controlled argon environment. The results revealed nonmonotonic frictional behavior with a higher running-in coefficient of friction (COF) followed by a lower steady-state COF, as a result of transition from two-body abrasion to three-body abrasion along with formation of a tribofilm. A key finding of this work is the sensitivity of the running-in COF to experimental conditions; maximum running-in values were lower at either elevated temperature (0.52–0.54) or reduced sliding speed (0.51–0.54). Conversely, the steady-state COF remained invariant at approximately 0.3 across all tested parameters. Transmission electron microscopy revealed a 0.5–2.0 μm thick nanocrystalline tribofilm that was thought to be formed by the compaction of the graphitic wear debris on the contact surface during the sliding process. The nanocrystalline nature of the tribofilm was further confirmed by Raman spectroscopy. As a result, the combination of tribological testing and morphological characterization provided a mechanistic understanding of the frictional behavior of nuclear graphite upon sliding.

Friction↗
DOE OSTI2011

Microstructural Characterization and Pore Structure Analysis of Nuclear Graphite

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

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Strength characterization of solar-synthesized cylindrical graphite carbon–carbon composites

The present work describes a new morphology of cylindrical graphite synthesized in a unique manner by direct solar decomposition of methane and reports the mechanical properties of its manifestation in woven carbon–carbon composites. Whereas traditional carbon–carbon composites are formed via epoxy impregnation, the deposited material in the process presented here conformally follows the existing shape and orientation of individual carbon fibers. Through this process, cylindrical graphite fibers are synthesized that possess superior strength due to their graphitic layers that amalgamate into interlocked pathways between disparate fibers. Here, strength measurements, taken in tandem with Raman, XRD, and SEM, paint a picture of the shift from cloth to composite behavior via graphitization of the original substrate combined with sheets of graphene coalescing into a unified composite, with a notable improvement in elastic modulus from 0.19 to 2.66 GPa.

Carbon–carbon composite↗
NASA NTRS1983

Formulation and Characterization of Epoxy Resin Copolymer for Graphite Composites

Maximum char yield was obtained with a copolymer containing 25% mol fraction DGEBE and 75% mol fraction DGEBA (Epon 828). To achieve the high values (above 40%), a large quantity of catalyst (trimethoxyboroxine) was necessary. Although a graphite laminate 1/8" thick was successfully fabricated, the limited life of the catalyzed epoxy copolymer system precludes commercial application. Char yields of 45% can be achieved with phenolic cured epoxy systems as indicated by data generated under NAS2-10207 contract. A graphite laminate using this type of resin system was fabricated for comparison purposes. The resultant laminate was easier to process and because the graphite prepreg is more stable, the fabrication process could readily be adapted to commercial applications.

Keck, F. L.↗
NASA NTRS1983

Mechanical Property Characterization of F-100/934 Graphite-Epoxy Composite Material

Procedures for quality control, fabrication techniques, specimen machining, test methodologies, and data collection and interpretation were defined using Thornel P-100 carbon fiber combined with Fiberite's 934 epoxy resin in unidirectional tape form. Tests on the prepreg material, including gel time, resin solids content, tack, volatile content, and flow are detailed. The steps of the fabrication and machining processes are described. Both destructive and nondestructive techniques were employed to assess the properties of the cured laminates. Tests to determine tensile strength and modulus, compressive strength and modulus, interlaminar and inplane shear strength, and flexural strength and modulus are discussed. The results of these experiments are compared with data generated by other researchers.

Clinton, R. G., Jr.↗
DOE OSTI2016

Property evaluation and microstructure characterization of the A3-3 matrix graphite - 18440

Spherical fuel elements with TRISO fuel will be used in the pebble-bed high temperature gas-cooled nuclear reactor. The fuel will be located within the center region of the elements, and will be suspended in, and surrounded by a shell of matrix graphite. The matrix graphite provides structural support to the fuel, prevents fuel damage, and allows for heat transfer from the fuel to the coolant. The raw materials for the matrix graphite, type A3-3, includes 64 wt% natural flake graphite, 16 wt% artificial graphite and 20 wt% phenol resin binder. After the final heat treatment, the resulting A3-3 matrix is composed of 71 wt% natural graphite, 18 wt% artificial graphite, and 11 wt% phenol resin-derived coke carbon. The microstructure of A3-3 matrix graphite has direct influences on the material properties, which are crucial to maintaining the integrity and safety of the fuel elements. This presentation will discuss microstructure features, such as lattice, and crystal parameters, porosity, pore size and distribution, and anisotropy. The multiple physical, mechanical and thermal properties of this material, including: elastic modulus, mechanical strength, and oxidation property will also be presented. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Graphitic carbon nitride/Cu{sub 2}O heterojunctions: Preparation, characterization, and enhanced photocatalytic activity under visible light

As a metal-free semiconductor material, graphitic carbon nitride (C{sub 3}N{sub 4}), the high recombination rate of photogenerated charges and insufficient sunlight absorption limit its solar-based photocatalytic activity. Here, we reported the heterojunctions of C{sub 3}N{sub 4}–Cu{sub 2}O with a p–n junction structure, which was synthesized by a hydrothermal method. The HR-TEM result revealed an intimate interface between C{sub 3}N{sub 4} and Cu{sub 2}O in the heterojunction, and UV–vis diffuse reflection spectra showed their extended spectral response in the visible region compared with pure C{sub 3}N{sub 4}. These excellent structural and spectral properties, as well as p–n junction structures, endowed the C{sub 3}N{sub 4}–Cu{sub 2}O heterojunctions with enhanced photocatalytic activities. The possible photocatalytic mechanism that photogenerated holes as the mainly oxidant species in photocatalysis was proposed base on the trapping experiments. - Highlights: • A hydrothermal method was used to prepare C3N{sub 4}–Cu{sub 2}O heterojunction. • The resulting heterojunction possesses broader absorption in the visible region. • The material owns a high visible light activity and stability for dye degradation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
NASA NTRS1984

Space Radiation Effects on Graphite-Epoxy Composite Materials

Radiation effects on engineering properties, dimensional stability, and chemistry on state of the art composite systems were characterized. T300/934 graphite-epoxy composite was subjected to 1.0 MeV electron radiation for a total dose of 1.0 x 10(10) rads at a rate of 5.0 x 10(7) rads/hour. This simulates a worst case exposure equivalent to 30 years in space. Mechanical testing was performed on he 4-ply unidirectional laminates over the temperature range of -250 F (116K) to +250 F (394K). A complete set of in-plane tensile elastic and strength properties were obtained (E sub 1, E sub 2, nu sub 12, G sub 12, X sub T, Y sub T, and S). In addition electron microscopy was used to study and analyze the fracture surfaces of all specimens tested. Results indicate that little difference in properties is noted at room temperature, but significant differences are observed at both low and elevated temperatures.

Milkovich, S. M.↗
DOE OSTI2026

Linking DSC/TGA to Cell Levels: Energetics, Evolved Gases, and Thermal Safety of NMC811‐Graphite Micro‐Cell

Thermochemical characterization of battery materials links intrinsic material properties to decomposition pathways, heat generation, and gas evolution that govern performance and safety. Despite extensive work on NMC811-Graphite, variability across partial configurations and the limited adoption of micro-cell architectures (cathode+anode+electrolyte+separator) hinder robust cell-scale interpretation. Accordingly, this work establishes a bottom-up, component-resolved methodology integrating DSC/TGA, evolved gas analysis (EGA), and in situ XRD to link decomposition pathways and energy release across partial and micro-cell configurations, providing a transferable assessment of safety and stability in emerging chemistries. In separator-free configurations, the gas–solid reaction between cathode-evolved O 2 and anode-leached Li dominates the net heat release (1139 J g −1 ). In contrast, in the micro-cell configuration, the separator hinders O 2 transport and alters the timing and pathways of other reactions, and reduces the net energy release to 618 J g −1 . Energy release was organized into defined temperature windows that provide a framework for a thermodynamic model combining quantified gas evolution with selected decomposition pathways and effective reaction enthalpies to estimate net specific energy release, with agreement between DSC and cell-level tests. Ex situ XPS of heat-treated samples extends post-mortem analysis to thermal-abuse regimes, supporting key pathway elements.

25 ENERGY STORAGE↗
DOE OSTI2019

Comparative analysis of microstructure and reactive sites for nuclear graphite IG-110 and graphite matrix A3

Nuclear graphite and graphite matrix are both high-purity, fine-grain, graphitized materials designed for use in Fluoride-Salt Cooled High Temperature Reactors (FHRs), High Temperature Gas Cooled Reactors (HTGRs), and Molten Salt Reactors (MSRs). Their ability to chemisorb radioisotopes (such as tritium) and their chemical stability is dependent upon the density of reactive carbon sites (RCS), which varies among grades of graphite and changes with neutron irradiation. While nuclear graphites have been extensively characterized, much less data is available for graphite matrix. The current study performs a comparative analysis of IG-110 nuclear graphite and A3 graphite matrix, by X-ray Diffraction (XRD), Raman, and H 2 uptake at 700 °C with H 2 partial pressure range of 300 Pa and 22 kPa. At 10 kPa, the RCS occupied by H 2 uptake per number of carbon atoms at the edge of a crystallite, H/C edge , is 600(300) appm for IG-110 and 2300(200) appm for A3. Prior empirical studies for nuclear graphite establish a linear relationship between H/C and C edge /C and degree of graphitization. The present study shows that these correlations no longer hold when extrapolating to graphite matrix; the type of defects and the defect density at the crystallite surface are different between IG-110 and A3 due to differences in raw materials and manufacturing method.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

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

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-14 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)

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Synthesis of Nano-Crystalline Graphite Films in Hollow Cathode Discharge

The results of experiments on the production of nano-crystalline graphite films in the hollow cathode discharge on Ni(111)/sapphire substrate are presented. The characteristics of the discharge for different gases and pressures are given. The emission spectra of the plasma have been measured. The layers of the deposited graphite have been characterized by the methods of Raman scattering and atomic force microscopy. The field emission of nano-crystalline graphite was also measured. The produced layers have good homogeneity and high value of the electron field emission. The presence of vertically growing graphene is revealed.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2016

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

As part of an overall effort to convert US research reactors to low-enriched uranium (LEU) fuel, 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 UO{sub 2} 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 characterization of the graphite and phenolic resin based fuel matrix. Methods included Scanning Electron Microscopy, Raman spectroscopy, and X-ray Diffraction. 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 potentially be obtained. Although no one type of analysis has provided a complete determination of the quantities and identities of all carbon phases present in the fuel matrix, additional information has been gathered and is being used to refine future analysis methods. From the three types of quantification performed, results have been within 1% (92.8%, 93.02% and 93.6% for BSE image analysis, XRD and Raman spectroscopy, respectively). More work remains to identify any trace crystalline carbonaceous phases that may be present, quantify any traces that are identified, and develop and use further complementary bulk phase quantification techniques. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

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

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 characterization 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.

10 SYNTHETIC FUELS↗
DOE OSTI2020

Electrochemical behavior and separation of iodide in molten LiF-NaF-KF melt

This work studied the electrochemical behavior of iodide in molten LiF-NaF-KF mixture and the separation of iodide from the molten salt at 973 K using a graphite working electrode. The cyclic voltammetry (CV) and chronoamperometry (CA) measurements indicated an anodic passivation layer was formed on the graphite electrode in this melt and the layer blocked the charge transfer across the electrolyte/electrode interface. However, the added I- ions could pass through the passivation layer and were oxidized at the graphite electrode. The characterization of the obtained products proved that iodide ions was efficiently oxidized to I2 gas during electrolysis. Furthermore, the graphite anode showed excellent stability without any obvious change in dimensions after electrolysis, which showed the high current efficiency of I2 evolution reaction. Thus, the graphite electrode was possible to be served as an excellent anode to separate other anions (such as Br) from molten fluoride salts.

36 MATERIALS SCIENCE↗
DOE OSTI2016

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

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

36 MATERIALS SCIENCE↗
DOE OSTI2022

Multiscale characterization and comparison of historical and modern nuclear graphite grades

Beginning with Chicago Pile I, graphite has been used as a moderator material in nuclear power stations and is considered a potential material for use in future Generation IV advanced reactors. The microstructure of graphite is responsible for much of its mechanical and thermo-physical properties, and how it responds to irradiation. To understand graphite microstructure, it is necessary to understand its porosity at the macro- and micro-scales; and to understand its porosity, it is necessary to characterize the morphological connectivity of the void content and the two main phases of graphite: filler and binder. Here, using several microscopy and analytical techniques, a detailed examination of the heterogeneity, microstructure and pore structure of different graphite grades and their binder and filler phases is presented. Significant differences were found between coarser and finer nuclear grades. Coarse grades have a more diverse range of filler particles, pores and thermal cracks. Finer grades have a more well-defined pore size distribution, fewer variations of filler particles sizes and do not contain as many large thermal cracks. Fine grades tend to have a well-connected network of pores whereas coarser grades contain a larger content of closed porosity. The framework developed within this work can be applied and used to assess the various graphite grades that would down-select materials for specific use in graphite moderated reactor designs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗