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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
NASA NTRS2011

Stable dispersions of polymer-coated graphitic nanoplatelets

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.

Stankovich, Sasha↗
NASA NTRS1956

Energy of Cohesion, Compressibility, and the Potential Energy Functions of the Graphite System

The lattice summations of the potential energy of importance in the graphite system have been computed by direct summation assuming a Lennard‐Jones 6–12 potential between carbon atoms. From these summations, potential energy curves were constructed for interactions between a carbon atom and a graphite monolayer, between a carbon atom and a graphite surface, between a graphite monolayer and a semi‐infinite graphite crystal and between two graphite semi‐infinite crystals. Using these curves, the equilibrium distance between two isolated physically interacting carbon atoms was found to be 2.70 a, where a is the carbon‐carbon distance in a graphite sheet. The distance between a surface plane and the rest of the crystal was found to be 1.7% greater than the interlayer spacing. Theoretical values of the energy of cohesion and the compressibility were calculated from the potential curve for the interaction between two semi‐infinite crystals. They were ΔE c = —330 ergs/cm 2 and β = 3.18×10 —12 cm 2 /dyne, respectively. These compared favorably with the experimental values of ΔE c = —260 ergs/cm 2 and β = 2.97×10 —2 cm 2 /dyne.

L A Girifalco↗
NASA NTRS2021

Tunable Pseudocapacitive Intercalation of Chloroaluminate Anions into Graphite Electrodes for Rechargeable Aluminum Batteries

Rechargeable aluminum-graphite batteries using chloroaluminate-containing electrolytes have been the focus of significant research, particularly due to their high-rate capabilities. Engineered graphite electrodes have been shown to exhibit supercapacitor-like rate performance, despite the fact they store charge via the electrochemical intercalation of polyatomic AlCl4− anions. However, the origins of such rate capabilities are not well understood. Here, using electrochemical techniques, we disentangle quantitatively the diffusion-limited Faradaic, pseudocapacitive, and capacitive contributions to charge storage, revealing that AlCl4− anions intercalate into graphite with significant pseudocapacitive characteristics due to low ion diffusion limitations. Pristine and mildly exfoliated graphites are compared, where exfoliation resulted in significantly higher pseudocapacitive AlCl4− intercalation at the highest potential redox pair as well as higher galvanostatic capacity retention at faster discharge rates. The relationships between graphite structure, ion mass transport, and the overall rate of electrochemical AlCl4− intercalation are discussed. Ion diffusion within the electrolyte phase of the porous electrode is shown to play a key role in controlling the rate of intercalation at higher potentials and faster rates, which can be enhanced by reducing electrode tortuosity. The results establish that chloroaluminate anion intercalation into graphite exhibits non-diffusion-limited pseudocapacitive contributions that are tunable by modifying the graphite structure.

Jeffrey H. Xu↗
DOE OSTI2020

Radiation Effects in Graphite

Graphite has been a core structural material and neutron moderator since the first man-made chain reaction experiment in Chicago. This article will discuss the requirements of what makes graphite a good solid moderator. The manufacture and properties of some currently available near-isotropic and isotropic grades are described. The major features of a graphite moderated reactors are briefly outlined. Displacement damage and the induced structural and dimensional changes in graphite are described. Recent characterization work on nano-carbons and oriented pyrolytic graphites that have shed new light on graphite defect structures are reviewed, and the effect of irradiation temperature on the defect structures is highlighted. Changes in the physical properties of nuclear graphite caused by neutron irradiation are reported. Finally, the importance of irradiation induced creep is presented, along with current models and their deficiencies.

Campbell, Anne↗
DOE OSTI2021

Matrix Graphite Material Models In Pebbles and Compacts For Bison

The cores and reflectors in high-temperature gas-cooled reactors (HTGRs) are made of graphite materials, with the graphite acting as a moderator, a fuel host matrix, or the foundation for various structural components. This study aims to survey the models in the literature for graphite materials being used as host matrices in pebble/fuel compacts and to implement those surveyed models into Bison to conduct an early assessment of graphite's thermo-mechanical response under various reactor conditions. In this study, thermal (e.g., thermal conductivity, and specific heat capacity) and mechanical (e.g., elastic properties, thermal expansion, irradiation-induced dimensional changes, and irradiation-induced creep) material models for various graphite grades (e.g., H-451, IG-110, G-348, 2020, A3-3, and A3-27) are incorporated into Bison. Two benchmark problems are then exercised utilizing these new graphite-related capabilities: (1) modeling an Advanced Gas Reactor (AGR)-2 fuel compact, and (2) modeling the debonding of a particle-matrix interface.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Options, Initial Design Requirements, Estimated Costs, Reactor Commitments, and Potential Uses of a Graphite Leadout Type Experiment Supporting Various Commercial HTR Vendors

Multiple commercial High Temperature Reactor (HTR) vendors and nuclear graphite suppliers would benefit by collaborating on a new irradiation capsule(s) that would include graphite grades not included within the AGC Experiment. This new irradiation capsule(s) would provide data to answer vendor graphite licensing issues. Rather than spending money (and especially) time in designing separate irradiation capsules for each designer, the capsule(s) would be used for multiple graphite and composite designs to maximize efficiency and promote multiple HTR designs. However, the primary motivation for assisting vendors with this new irradiation capsule(s) is lack of availability in the existing Material Test Reactors (MTRs). Cost reduction is a secondary goal. A common, collaborative, capsule design can be achieved for graphite and composites due to the similarity of different grades. Irradiation, disassembly, shipping, and PIE costs would be cost-shared by all users. It is anticipated that interest would extend across all DOE campaigns (micro-Rx, SMR, GCR, MSR, etc.) due to the similar requirements for all graphite grades.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Sustainable Graphite and Jet Fuel from Biorefinery Residue

Abstract Battery‐grade graphite and aviation fuel are traditionally produced from non‐renewable, fossil carbon feedstocks and result in substantial greenhouse gas emissions. Biomass holds exciting potential as a renewable and sustainable feedstock for the production of graphite and aviation fuel, but challenges exist including the necessity of a catalyst when producing graphite and low selectivity when producing aviation fuel. A process to convert a biomass‐derived feedstock into graphite without the use of a catalyst and fuels with high selectivity towards sustainable aviation fuel (SAF) is innovated. Heavy bio‐oil undergoes a conversion process similar to the commercial production of synthetic graphite including coking at 500 °C, calcination at 1000 °C, and graphitization at 2800 °C. The resulting biographite exhibits excellent performance in lithium‐ion battery configurations with specific capacity of ~330 mAh g −1 and a 96.8 % capacity rebound after high rate cycling. The liquid hydrocarbon co‐product from coking is suitable for hydrotreating into SAF. The aviation fuel fraction (70 wt % of the fuel produced) meets ASTM standards and is composed primarily of cycloalkanes (~80 wt %) which improves energy density compared to paraffins produced by other SAF pathways and may replace aromatics for elastomer swelling in traditional jet fuel with less soot production.

09 BIOMASS FUELS↗
DOE OSTI2019

High performance graphene-like thinly layered graphite based visible light photodetector

In this work, the drop casting method is utilized to obtain a thin graphene-like graphite layer as a photodetection layer on a SiO{sub 2}/Si substrate, and its performance is characterized to determine its potential use in commercial applications. The thin layer is formed by first synthesizing graphite soot from a butane flame before being turned into a liquid solution and drop-casted onto the substrate to form a thin graphene-like graphite film. Field emission scanning electron microscope and Raman vibrational mode analyses indicate that the fabricated graphene-like graphite layer has characteristics similar to the graphene. The graphene-like graphite photodetector demonstrates a narrow photoresponse from 530 to 680 nm, covering visible sources from green to red. The responsivity and external quantum efficiency of the device under the illumination of red laser 660 nm is found to be around 148 mA W{sup −1} and 27.8% respectively and is faster than that reported for similar systems using graphene and reduced graphene oxide previously. A fast response time of 83.7 and 28 µs at a modulation frequency of 1.0 and 10000 Hz respectively from the graphene-like thinly layered graphite photodetector shows potential application for the development of low-cost carbon-based photodetectors in near future.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2021

Neutron thermalization in nuclear graphite: A modern story of a classic moderator

Graphite has been of great interest since the dawn of the nuclear era. In this work, we will focus on the thermal neutron scattering of two types of nuclear graphite that have different microstructures, namely, the historic PGA which has a very anisotropic medium-grained needle coke, and the new G347A which is a near-isotropic fine-grained graphite. We report the results of room temperature inelastic neutron scattering measurements for the scattering function, double differential scattering cross-section, and phonon density of states. The obtained results are discussed in conjunction with results extracted from first-principles lattice dynamics simulations of crystalline graphite, using the incoherent approximation and by also including the coherent-one phonon contribution. Excellent agreements have been found between the measured and calculated scattering functions and phonon densities of states. The inclusion of the coherent-one phonon scattering shows a significant improvement of the calculated scattering functions, double differentials, and inelastic scattering cross-sections. We also highlight the physical concerns about the current thermal neutron scattering ENDF/B-VIII of nuclear graphite library and its corresponding phonon density of states. This work aims to provide a new presentation of neutron thermalization by graphite and offers a new set of high quality experimental and computational data for consideration.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Electron tomography of unirradiated and irradiated nuclear graphite

Graphite is the moderator material of several Generation IV nuclear reactor concepts, as well as the British Advanced Gas-cooled Reactors (AGR). Porosity can heavily influence the material properties, mechanical irradiation response, and neutron induced shrinkage or swelling of nuclear-grade graphite. Due to the sub-micron size of several types of pores found in graphite, only a high-resolution imaging technique such as electron tomography are capable of visualizing these features in three dimensions. In this research, we used electron tomography to characterize as-received and neutron irradiated samples of IG-110 nuclear-grade graphite to show for the first time the 3D structure of both native and irradiation-induced nano-cracks. This technique also reveals unique characteristics of graphite such as the structure that surrounds pores and could be used to inform molecular dynamic simulations of irradiated graphite and experimental techniques such as gas-absorption. This work also shows the utility of this technique for the study of other nuclear porous carbon-based materials.

36 MATERIALS SCIENCE↗
DOE OSTI2023

High temperature annealing of irradiated nuclear grade graphite

Previous work has shown that the material properties of nuclear grade graphite are substantially affected by the atomic and microstructural changes that occur during neutron-irradiation. The parameters that play a role in these changes are irradiation dose, temperature, graphite composition and the initial microstructure of the graphite. Understanding the details of how these material changes occur and exactly what the changes consists of will enable the prediction of the material property changes as a function of these variables. Other work has shown these atomic level and microstructural changes can be healed or annealed out by raising the irradiated graphite above its irradiation temperature. Here in this work, experiments were carried out to investigate how the properties of irradiated graphite recover when heated. By showing property recovery as a function of annealing temperature or energy, insight is provided into the type of damage that occurred during neutron-irradiation. The data presented here shows recovery of thermal diffusivity, coefficient of thermal expansion, Young's modulus and electrical resistivity between annealing temperatures of 500 °C and 2380 °C. Graphite grades NBG-18, IG-110 and PCEA are considered that were both stressed and unstressed during irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Microstructural changes in nuclear graphite induced by thermal annealing

Irradiation-induced property change in nuclear graphite is particularly important when considering the in-service lifetimes of graphite components. In other works, it has been shown that annealing of irradiated graphite above the irradiation temperature may heal atomic-level defects and thus reverse some amount of physical property change. In this work, virgin nuclear graphite IG-110 was annealed at 2500°C to observe any microstructural changes due solely to high-temperature thermal annealing. We report the results shown in this study suggest that fullerene-like defects will arise due to thermal annealing near graphitization temperature. Consequently, such defects may directly contribute to non-recoverable physical property change which has previously been observed in irradiated nuclear graphites.

36 MATERIALS SCIENCE↗
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 OSTI2024

Synthesizing Highly Crystalline Graphite Powder from Bulk Polyethylene Waste for Lithium-Ion Battery Anodes

Upcycling plastic waste into graphite can potentially be used, in conjunction with other methods, to manage existing waste materials and diversify graphite supply chains. However, synthesizing large quantities of crystalline graphite powder from plastic waste, particularly polyethylene (PE), remains a challenge because PE decomposes into light gases during thermal processing and simple methods do not exist at any appreciable size scale to address this challenge. In this work, a method is developed for air processing bulk forms of PE waste to create stable carbon char that does not readily decompose during high-temperature processing. This method employs solid additives in the form of salts, which are combined with the PE melt during air processing to increase the effective surface area of the melt and improve the oxygen-driven chemistry that stabilizes PE for high-temperature processing. After removal of the solid salt additives from the PE-derived char, it is converted into a highly crystalline bulk graphite powder using an Fe-based catalytic process. The PE-derived graphite anode in a lithium-ion coin cell showed a specific capacity of 345 mAh/g at 0.05C with an initial Coulombic efficiency of 87% and reversible capacity retention of ~100% at different current rates. It also showed a specific capacity of up to 313 mAh/g at 0.5 discharge/charge cycles per hour (0.5C) and Coulombic efficiency of 99.9% after 250 cycles, indicating excellent electrochemical performance as an anode material for lithium-ion batteries. This method illustrates that there are opportunities for upcycling large quantities of PE waste to produce graphite powders suitable for use in LIBs.

25 ENERGY STORAGE↗
DOE OSTI2024

Solvent-Mediated, Reversible Ternary Graphite Intercalation Compounds for Extreme-Condition Li-Ion Batteries

Traditional Li-ion intercalation chemistry into graphite anode exclusively utilizes the co-intercalation-free or co-intercalation mechanism. The latter mechanism is based on ternary graphite intercalation compounds (t-GICs), where glyme solvents were explored and proved to deliver unsatisfied cyclability in LIBs. Herein, we report a novel intercalation mechanism, that is, in-situ synthesis of t-THF-GICs in the tetrahydrofuran (THF) electrolyte via a spontaneous, controllable reaction between binary-GICs and free THF molecules during initial graphite lithiation. The spontaneous transformation from b-GIC to t-GIC, which is different from conventional co-intercalation chemistry, is characterized and quantified via operando synchrotron X-ray and electrochemical analyses. The resulting t-GIC chemistry obviates the necessity for complete Li-ion desolvation, facilitating rapid kinetics and synchronous charge/discharge of graphite particles even under high current densities. Consequently, the graphite anode demonstrates unprecedented fast charging (1 min), dendrite-free low-temperature performance, and ultralong lifetimes exceeding 10,000 cycles. Full cells coupled with layered cathode, display remarkable cycling stability upon a 15-min charging and excellent rate capability even at -40 °C. Furthermore, our chemical strategies are shown to extend beyond Li-ion batteries to encompass Na-ion and K-ion batteries, underscoring their broad applicability. Our work contributes to the advancement of graphite intercalation chemistry and presents a low-cost, adaptable approach to achieving fast-charging and low-temperature batteries.

25 ENERGY STORAGE↗
DOE OSTI2019

The Degradation of Strength under Varying Oxidizing Conditions for Nuclear Graphite

Strength behavior of fine and medium grain size nuclear graphite grades are presented after being subjected to varying oxidation conditions. This study provides valuable information for design and eventual licensing considerations for the oxidation behavior of graphite components within a gas-cooled High Temperature Reactor design and the eventual structural integrity of graphite core components. These tests address the underlying oxidized strength of nuclear graphite components for oxidized mass losses up to and beyond the current limits recommended for graphite components in the ASME code (maximum mass loss = 10%) over a range of oxidation temperatures (550° C – 750° C). Preliminary results generally demonstrate that low temperature oxidation can result in higher strength reductions than oxidation at higher temperatures for similar mass loss levels. Since the graphite component strength is a critical parameter necessary to determine the core structural integrity, determining the actual residual strength after oxidation is essential for determining actual component oxidation limits within the ASME code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Initial Fracture Propagation Modeling of Graphite Components with Grizzly

Graphite has historically been extensively used in power reactor cores and will be used in multiple types of advanced reactors currently under development. These graphite structural components can experience significant stresses due to nonuniform volumetric strains induced by irradiation and thermal expansion, which can lead to fracture. Robust tools for predicting fracture initiation and propagation in graphite structural components in nuclear reactors are important for evaluating component integrity, developing design standards, and interpreting experimental results to characterize graphite performance. The U.S. Department of Energy’s Nuclear Energy Advanced Modeling and Simulation program has been developing degradation models for other structural components in nuclear reactors within the Grizzly and BlackBear codes. This report documents an effort to develop initial capabilities for modeling graphite fracture within these codes, building on prior efforts to model fracture in other materials. Major elements of this effort include developing a new system for modeling fracture nucleation and growth in two dimensions using the extended finite element method and incorporating a damage and plasticity model. These capabilities are applied here to model a representative graphite component and a splitting disc experiment used to obtain tensile strength.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Perspective on “code qualifying” new graphite grades for use in advanced nuclear reactors*

The American Society of Mechanical Engineers (ASME) publishes the Boiler and Pressure Vessel (BPV) Code, which include guidance for the safe development, construction, and operation of boilers and pressure vessels. ASME BPV Code Section III “Rules for Construction of Nuclear Facility Components” Division 5 focuses on “High Temperature Reactors”. Subsection HH, subpart A lists the different materials properties that are to be measured and how those properties change due to different environmental conditions (oxidation and irradiation damage) for a graphite to be accepted for use in a high temperature reactor core (i.e., “Code Qualified”). Currently there are no nuclear graphite grades that are “Code Qualified” (i.e., a reactor designer can select a graphite grade and build their reactor without any additional testing), which is due in part to development of new graphite grades in the last 20 years and the lack of comprehensive programs needed to produce the data for the code cases. This perspective is going to discuss the requirements, as called out in the ASME BPV Code, that are necessary to “code qualify” a nuclear graphite grade but will primarily focus on the practical and technical challenges associated with irradiation-induced property changes and how to address these to assist with getting graphite ready for use in advanced nuclear reactors. These same technical challenges can be expected to arise for other materials being developed for advanced reactor concepts.

advanced nuclear reactors↗