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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Review of Graphite Properties Relevant to Micro-Reactor Design

Microreactors are advanced reactors that are designed to be small, scalable, and portable with up to 20 MW power. These reactors could operate independently, or as part of a microgrid, without need of a frequent refueling. Micro-reactors are designed to be deployable to provide electric power in remote locations, therefore providing greater flexibility. Several micro-reactors are under development in the US. These reactors are based on advanced gas-cooled or heat-pipe concepts. While Graphite can be found in various applications, some high-temperature reactors plan to employ Graphite in their core components due to its role as a moderator and its ability to maintain its structural integrity when subjected to elevated temperatures and high neutron fluence. Although there is significant experience with use of Graphite in nuclear applications, there are still unknowns that should be addressed. Graphite properties depend on the graphite type and affected by the fabrication procedures. Material properties are functions of the temperature and irradiation dose. The fundamental properties also affected by the oxidation environment. Therefore, in-depth understanding of graphite properties, including the effect of irradiation, oxidation and other influencing parameters are required to evaluate the design accurately. This draft aims to provide a comprehensive review of material properties for Graphite relevant to micro reactor design. For this work, two distinct grades of graphite were considered, namely IG-110 and Mersen-2114. Physical, thermal, and mechanical properties for these grades were comparatively evaluated. The properties that are discussed in this draft aim to serve as a basis for design and safety analysis.

42 ENGINEERING↗
DOE OSTI2021

Compressed Expanded Natural Graphite (CENG) Processing for PCM Composites

The use of phase change materials (PCMs) in thermal energy storage applications has received considerable attention in recent decades. Organic PCMs are popular due to their high latent heat of fusion, noncorrosive properties, and relative stability over many charge and discharge cycles. A primary limitation of these materials is their low thermal conductivity. This has led researchers to develop various methods to increase thermal conductivity by seeding PCM with or impregnating them into more conductive materials. One method is to impregnate PCMs into compressed expanded natural graphite (CENG) matrices, which can improve thermal conductivity by a factor of 100. CENG matrices have received particular interest due to their low cost, high porosity, small (nano/micro) pore size, high pore density, high thermal conductivity, and ability to be compressed into many geometries. PCM/CENG matrix composites have been extensively studied; however, the effect that CENG processing has on PCM saturation and the overall matrix thermal conductivity has not been well investigated. This processing includes four major steps including graphite intercalation, thermal shock, compression, and PCM saturation. Intercalation involves soaking graphite flakes in sulfuric and/or nitric acid to intercalate the acids between the graphene layers. The graphite flakes are then subjected to a high-temperature thermal shock, during which the intercalated acid is gasified rapidly, pushing the graphene layers apart, resulting in accordion-shaped graphite "worms". The "worms" are then compacted to a desired bulk density and then soaked with molten PCM until fully saturated. The properties of the produced CENG matrix, and its ability to allow PCM permeation, are sensitive to the processing parameters, namely, the thermal shock temperature and exposure time, as well as the matrix apparent density or porosity. Here, we study the effect of the thermal shock conditions necessary to expand intercalated graphite flakes on PCM saturation and the expanded graphite's thermal conductivity and morphology. We found that the thermal shock temperature exhibits the greatest influence. At greater shock temperatures, SEM images showed that expanded graphite worms exhibited greater density of pores, thus increasing total surface area within the matrices. Increasing thermal shock temperature yielded greater overall PCM saturation, as well as an increased rate of saturation. Improvements in PCM saturation rate and overall saturation are obtained as the shock temperature is increased. Longer exposure to thermal shock also improves initial saturation rates and is beneficial if a shortened impregnation time is needed. Thermal shock conditions did not impact thermal conductivity; however, conductivity was largely affected by matrix porosity. A local maximum in axial thermal conductivity was observed at around 83% porosity, which is similar to that observed in previous studies.

compressed expanded natural graphite↗
DOE OSTI2024

ASME Design Code Rule Changes for Nuclear Graphite

The American Society of Mechanical Engineers Boiler Pressure and Vessel Code (ASME BPVC) Section III, Division 5, Article HHA-3000 outlines graphite core component and graphite core assembly design guidelines. Graphite core components are defined as ?components manufactured from graphite that are installed to form a graphite core assembly within the reactor pressure vessel of a high temperature, graphite moderated fission reactor.? (p. 413) Graphites? inherent defect distributions do not allow for deterministic material reliability. Rather, graphite has variable strength distributions which change by grade. Article HHA-3000 outlines two semi-probabilistic methods, the full and simplified assessments, which set design load limit targets for each of three component structural reliability classes. The Design Task Group was officially recognized as a specialized task group within ASME November of 2023, though we?ve been collaborating since 2022. The purpose of the Design Task Group is to correct, clarify, and make HHA-3000 function as intended. The Design Task Group will sunset once we?ve achieved our objectives. The Design Task Group was specifically told to not write new Code. While there may be more precise and more accurate methods to determine reliability targets, the current methods are conservative, relatively simple to implement, and have thus far been considered satisfactory for setting design reliability targets. Much of the ground-work to write proposal files and background documents for records to make the changes needed to achieve our objective have been completed. The Design Task Group has documented much of their work through papers, presentations, and memorandums. Three memorandums in which INL team members had substantial contributions are found in the Appendices: FEA Modeling for the Baseline Program, Evaluating the Effects on Margin of Updating the Threshold and Shape Parameters in the Full Assessment, and Interpretations of the Full and Simplified Assessments in ASME BPVC. Most of the on-going work to achieve the Design Task Group?s objective will be addressing comments on existing records and moving records through the balloting process. The Design Task Group met bi-weekly mostly through the end of FY2023. Since February 2024, the Design Task Group has mostly been completed with solving and documenting the technical issues associated with the assessments. Unless new tasks are identified, the remaining work of the Design Task Group will be political and editorial.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Nuclear Graphite

The basis for graphite in nuclear core components. Why graphite, what makes it tick, what is it's behavior, and some speculation on graphite behavior. How does graphite fit into the discussion for different types of reactors and what types of reactors are considering graphite. The importance of graphite for future reactor designs and what work needs to be conducted in characterizing nuclear graphite.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

HDG-1 Graphite Preirradiation Data Package Report

This report documents all pre-irradiation examination material-property measurement data for graphite specimens that are going to be used within the first high dose graphite (HDG) -1 irradiation capsule. The two new HDG capsules signify a major change to the AGC Experiment. HDG-1 and HDG-2 will replace the last two Advanced Graphite Creep (AGC) capsules (AGC-5 and AGC-6) which were designed to irradiate graphite at the extreme upper operational temperatures for a very-high-temperature reactor (VHTR) design, 1100°C. These very high temperature AGC-5 and AGC-6 capsules have been repurposed to re-irradiated specimens (from AGC-2, AGC-3, and AGC-4) at the lower temperatures of 600°C and 800°C. HDG-1 will be irradiated at 600°C and HDG-2 will be irradiated at 800°C. By re-irradiating the previous AGC specimens a total maximum neutron dose of around 15 dpa (displacements per atom) can be achieved for all major graphite grades at irradiation temperatures of 600°C and 800°C. Specimens in the HDG-1 capsule are made up of previously irradiated specimens from the AGC-2 capsule and unirradiated specimens prepared for the now discontinued AGC-5 capsule. Utilizing the irradiated specimens, a maximum neutron dose of around 15 dpa is anticipated. These new maximum dose levels will provide irradiated material property data over a total neutron dose range of 1-15 dpa at a temperature of 600°C when combined with the previous AGC-1 and AGC-2 irradiation data. This will provide quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades for use within high temperature reactor designs. Similar to previous AGC test trains, HDG-1 includes the major graphite grades (IG-110, NBG-17, NBG-18, PCEA, and 2114) as well as adding the very fine-grain grade IG-430 which is of interest to the Molten Salt Reactor (MSR) designs. Also new to the HDG-1 capsule are 90 smaller geometry specimens designated as pencil specimens. These specimens take up only one third the space of a standard creep size specimen. This increased number of specimens will enhance property measurement statistics because they will provide 3 times the control specimen data at a position that would otherwise only have a single measurement.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Infiltration of molten fluoride salts in graphite: Phenomenology and engineering considerations for reactor operations and waste disposal

The possibility and consequences of salt-infiltration in graphite must be evaluated for graphite used in molten salt reactors (MSRs) and fluoride-salt-cooled high-temperature reactors (FHRs), which can be subjected to salt pressures as high as 500 kPa. The volume of graphite porosity infiltrated by salt can be measured by direct infiltration and it can be predicted from the graphite pore size distribution, the surface tension of the salt, and the contact angle between the graphite and the salt. While these three properties are believed to be insensitive to irradiation, the former can be impacted by chronic or acute oxidation, and the latter two are highly sensitive to the chemistry of the salt and to events such as air ingress. For MSRs, predictions based on nominal properties of salt and graphite reveal that few graphite grades would satisfy the 4 vol% limit set in the Molten Salt Reactor Experiment, and even fewer would satisfy the 0.5 vol% design target. For FHRs, infiltration limits have not been defined and depend on the effect of infiltration on graphite properties, which are discussed. A hypothesis is presented for properties that may be impacted by infiltration and for which future studies are needed.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Sustainable conversion of biomass to rationally designed lithium-ion battery graphite

The carbon net negative conversion of bio-char, the low value byproduct of pyrolysis bio-oil production from biomass, to high value, very high purity, highly crystalline flake graphite agglomerates with rationally designed shape and size tailored for lithium-ion battery energy storage material is reported. The process is highly efficient, 0.41 g/Wh; the energy content of its co-product of the process, bio-oil, exceeds that needed to power the process. It is shown that the shape of the starting material is retained during the transformation, allowing the ultimate morphology of the graphite agglomerates to be engineered from relatively malleable biomass. In contrast to commercial graphite production, the process can be performed at small scale with low equipment costs, enabling individual research laboratories to produce Li-ion grade graphite with customizable shape, size and porosity for Si/graphite composite and other graphite involved anodes. The mechanism of the graphitization of bio-char, a “non-graphitizable” carbon, is explored, suggesting the molten metal catalyst is absorbed into the pore structure, transported through and transforming the largely immobile biochar. Finally, the transformation of biomass to rationally designed graphite morphologies with Li-ion anode performance that closely mimic commercial shaped graphite is demonstrated.

09 BIOMASS FUELS↗
DOE OSTI2020

Transient computational and experimental thermal analysis of graphite foam monoblock

A densified graphite foam is being explored for its applicability as plasma-facing material in fusion devices. Three different graphite foam monoblocks are constructed and tested in the Garching Large Divertor Sample Testing Facility. The monoblock samples consist of graphite foam press-fit to a single tube, graphite foam cubes brazed to a single, and graphite foam cubes press-fit to a single tube. The tube is composed of CuCrZr with a steel twisted tape. The monoblocks are exposed to the heat fluxes of 5, 6, and 8 MW/m 2 for 30 s to determine the maximum surface and body temperatures measured with thermocouple for each monoblock design. The press-fit monoblocks are exposed to 8 MW/m 2 for 15 s for 100 cycles to determine the effect of thermal cycling on the contact between the graphite foam and the tube. STAR-CCM+ is used to predict how much the contact between the foam and tubes varies as a result of thermal cycling. In addition, the 6 MW/m 2 loading is modeled in STAR-CCM+ to compare the transient cooldown curves of the computational results to the recorded temperatures at the surface and two different thermocouple locations. This comparison is used to validate the temperature-dependent thermal conductivity and specific heat capacity used in the models of the graphite foam. The computational modeling of the experiment has been used to hypothesize ways to use and improve upon the graphite foam as a suitable material for fusion applications.

Gehrig, Monica↗
DOE OSTI2022

Effects of Microstructural Composition, Porosity, and Microcracks on the Elastic Moduli of Nuclear Graphites

The elastic responses of nuclear graphite not only depend on the graphitic content itself but are largely dictated by the microstructural constitution of the material. The types of raw materials combined with the manufacturing processes used to produce the graphite yield the microstructural content that primarily includes graphite filler, graphitized pitch binder, and voids/defects that typically occupy approximately 20% of the volume. Among these microstructural components, porosity and microcracking (considered to be part of voids/defects) heavily influence the overall properties of the material, including the elastic moduli. It is widely accepted that the primary effect of oxidation is to increase porosity, but the related effect on the moduli cannot be explained satisfactorily by simply noting changes to porosity. Models describing the elastic moduli of porous, polycrystalline graphite materials have been developed to interpret experimental determinations of Young's modulus and shear modulus in oxidized graphites. Beyond porosity, the moduli are heavily influenced by microcracks, and their effects can be assessed 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. This work demonstrates how most modulus measurements in nuclear graphites can be understood using relatively simple models that describe the effects of porosity and microcracking. We also present directions to be pursued to account for microstructure-related effects that occur as a result of neutron irradiation.

Spicer, James B.↗
DOE OSTI2022

Understanding FLiNaK Salt Intrusion Behavior on Nuclear-Grade Graphite via Neutron Tomography

Graphite is an essential material as a neutron moderator in molten salt reactors (MSRs). To understand the impact of salt on the graphite structure to develop structural materials for MSRs, molten salt intrusion behavior on nuclear-grade graphite was studied. The graphite samples, IG-110 and PCEA, were tested for infiltration with LiF-NaF-KF (FLiNaK) at 750°C, 5 bar for 12 h, and, after the intrusion experiment, the graphite was analyzed by neutron imaging. The graphite and Li from FLiNaK showed great contrast in the neutron attenuation coefficient. Thus, the salt behavior in the graphite structure has been visualized for the first time without damaging the sample. The 3D image of the graphite was reconstructed after a neutron computed tomography scan, and the average salt coverage distribution of the XY surface in different depths was obtained from the reconstructed 3D images.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Report on initial development of a database of nuclear graphite characteristics based on microstructural characterization

This report outlines the current and future efforts to produce a comprehensive library of microstructures of nuclear graphite and carbon-based materials that are candidate materials for nuclear applications in the United States. This library must contain extensive characterizations of unirradiated graphite materials, a guide to some of the techniques used to characterize graphite, a compendium of characterization data of neutron-irradiated or oxidized material, and a compendium of microstructural information of carbon-based materials. These characterization efforts are being conducted at various length scales to understand these complex materials’ local structure and property relationships. Achieving this goal requires further developing or adapting advanced characterization techniques that capture graphite’s most relevant characteristics. Some of the general objectives of this project are to aid with the material selection, licensing, management, and core assessments of a graphite core by documenting the unirradiated microstructure of relevant grades or by characterizing the evolution of the microstructure under the reactor environment. Moreover, this project aims to provide additional information, guidelines for characterizing graphite, and a protocol to assess a nuclear graphite grade. This report also summarizes some of the initial results and some of the techniques commonly used to characterize nuclear graphite.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Design of Critical Experiments Involving Graphite elements at the IPEN/MB-01 Plate-Type Core

A series of critical experiments were designed involving the use of characterized graphite elements in the new plate-type core of the IPEN/MB-01 reactor. Eleven configurations with different graphite quantities are considered, each producing distinct effects on neutron fluxes and offering opportunities to analyze varying sensitivities to k eff in the system. The analysis shows that the experiments can be performed with acceptably low k eff uncertainties considering an adequate characterization of the graphite elements to be used. The sensitivity results obtained using SCALE TSUNAMI calculations show that the experiments have a notable k eff sensitivity coefficient to graphite. In the configurations tested, the k eff sensitivity to graphite TSL is low. This experiment will be useful for criticality safety validation of applications using uranium fuel enriched to around 19.75 wt % 235 U, light water and graphite. It will also be useful to gain more insight into graphite material properties and their effect on k eff as a result of using the well-characterized graphite elements. The execution of the experiments is planned for the summer of 2025. The 11 designed configurations serve as a basis for the final design, and fewer configurations will be executed. Once executed, these critical experiments will be evaluated and submitted for publication in the International Criticality Safety Benchmark Evaluation Project Handbook, supporting the DOE/NRC Collaboration for Criticality Safety Support for Commercial-Scale HALEU for Fuel Cycles and Transportation project goal.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Friction and wear of nuclear graphite exposed to molten FLiNaK salt environment

This report formally documents the completion of the Advanced Reactor Technologies Level 3 Milestone (M3TG-25OR0501104), “Initiate graphite wear studies of graphite samples exposed to molten salt environment,” due August 1, 2025. The report summarizes the ongoing activities aimed at characterizing the friction and wear behavior of graphite in molten LiF–NaF–KF (FLiNaK) salt environment. The wear and friction behavior of self-mated ET-10 nuclear graphite were tested at varying temperatures (550°C and 650°C), contact loads (40 and 80 N) and sliding speeds (1 and 10 mm/s) in a controlled argon environment. The results were compared to the wear and friction behavior of the same graphite material in a dry argon environment. This report presents initial studies on the friction and wear behavior of graphite in sliding contact mode, with future work focused on investigating the graphite matrix materials, as well as rolling and impact contact modes. The outcomes of this report could provide insights into assessing the integrity of the graphitic components of pebble-bed fluoride salt–cooled high-temperature reactors (PB-FHRs) to enhance their safe operation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Understanding the Reaction of Nuclear Graphite with Molecular Oxygen

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

J. Kane, Joshua↗
NASA NTRS1972

The ablation of graphitic materials in the sublimation regime.

A large variety of graphitic materials have been tested in an arc heated air stream at a surface pressure of 4.3 atm and a nominal surface temperature of 3925 K. Included were commercial and developmental grades of artificial graphites, both two and three dimensional carbon-carbon composites, composites seeded with refractory compounds, and several special materials such as pyrolytic graphite, mesophase graphite, glassy carbon, and natural graphite. ATJ graphite was used as a control material. Except for the seeded materials which had poor performance, the mass loss rate for all the man-made graphitic materials fell within the range of 17 per cent less to 30 per cent more than the rate for ATJ. Thus it is concluded that wide variations in constituents, processing, fabrication and structure have relatively little effect on the ablation performance of graphitic materials, at least under the conditions of the present tests. Particulate mass loss was observed for all the materials tested and is the dominant mechanism for mass removal at the present test conditions. It is suggested that this mechanism results from physical failure, primarily by compressive thermal stress.

Lundell, J. H.↗
NASA NTRS1981

Dynamics of solid dispersions in oil during the lubrication of point contacts. Part 1: Graphite

A Hertzian contact was lubricated with dispersed graphite in mineral oils under boundary lubrication conditions. The contact was optically observed under pure rolling, combined rolling and sliding, and pure sliding conditions. The contact was formed with a steel ball on the flat surface of a glass disk. Photomicrographs are presented which show the distribution of the graphite in and around the contact. Friction and surface damage are also shown for conditions when the base oils are used alone and when graphite is added to the base oils. Under pure rolling and combined rolling and sliding conditions, it is found that, for low speeds, a graphite film can form which will separate the contacting surfaces. Under pure sliding conditions, graphite accumulates at the inlet and sweeps around the contact, but very little of the graphite passes through the contact. The accumulated graphite appears to act as a barrier which reduces the supply of oil available to the contact for boundary lubrication. Friction data show no clear short term beneficial or detrimental effect caused by addition of graphite to the base oil. However, during pure sliding, more abrasion occurs on the polished balls lubricated with the dispersion than on those lubricated with the base oil alone. All observations were for the special case of a highly-polished ball on a glass surface and may not be applicable to other geometries and materials, or to rougher surfaces.

Cusano, C.↗
NASA NTRS1985

Meteoritic evidence that graphite is rare in the interstellar medium

Carbonaceous material from several distinct astrophysical environments that has been identified in primitive meteorites is discussed. Most of this material is not well crystallized graphite, but ranges from kerogenlike macromolecular organic matter to poorly graphitized carbon. Of the small fraction of the carbonaceous material which is graphite, most may have resulted from the graphitization of macromolecular precursors in the solar nebula. Since graphite is more stable than those precursors, preexisting interstellar graphite should have survived in meteorites to at least the same extent as did other forms of carbon. It is concluded from this dearth of graphite that graphitic carbon was not a major component of the interstellar dust at the time of the formation of the solar system.

Nuth, J. A.↗
DOE OSTI2021

Graphite-containing electrode and method related thereto

A graphite-containing electrode includes a porous body that has a plurality of first graphite-containing elements and a plurality of second graphite-containing elements intermingled with the first graphite-containing elements. The first graphite-containing elements have a first degree of graphitization and the second graphite-containing elements have a second, different degree of graphitization.

Darling, Robert Mason↗