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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Elevated Temperature Graphite Mechanical Testing

High purity graphite will be used for core components within most High Temperature Reactor (HTR) designs. Several "high tech” industries currently utilize synthetic, high-purity, commercially available graphite components to fabricate photovoltaic cells, semi-conductors, optical fibers, and other high value electronic industry products. New advanced HTR designs are also interested in using these graphite grades for long-term, internal core component applications. Consequently, the US Department of Energy, Advanced Reactor Technologies (DOE-ART) program has spent several years testing different high purity graphite grades for potential use within these new nuclear reactor designs. A significant part of that effort has been in the development and improvement of American Society for Testing and Materials (ASTM) test standards specifically for nuclear graphite grades. Nearly all ASTM test standards have either been developed or improved by the DOE-ART program over the past 25 years. The ART program continues to assist in the development of new ASTM test standards in support of the future commercial HTR fleet currently being designed and built in the USA. The newest effort undertaken by ART is the development of high temperature mechanical testing practices that may be acceptable for a future ASTM test standard (or guide) for this critical material property measurement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Materials property changes in ETU-10 graphite due to neutron irradiation at elevated temperatures

Graphite grade ETU-10, from IBIDEN Co., Ltd. Has been irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The irradiation program was developed to provide a preliminary study the irradiation-induced property changes to the dimensions/volume, elastic properties, strength, electrical resistivity, coefficient of thermal expansion, and thermal diffusivity/conductivity over a range of temperatures and neutron exposures that may be relevant for future nuclear reactors. The irradiation envelope covers a range of irradiation temperatures (300°C–900 °C) and fluences (up to 40 × 10 25 n/m 2 [E > 0.1 MeV] or ~30 dpa) that would be relevant for advanced nuclear reactors. Further, the dimensional change was observed to be anisotropic for an isotropic graphite, the specimen dimensions, volume, Young's modulus, shear modulus, and strength all displayed a parabolic fluence dependence, the electrical resistivity had a rapid rise followed by a decrease and a later increase, at high fluence the mean coefficient of thermal expansions was similar for all irradiation temperatures, and thermal conductivity rapidly decreased followed by a continued loss.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Damage tolerance of nuclear graphite at elevated temperatures

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 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.

36 MATERIALS SCIENCE↗
DOE OSTI2023

SAM Code Development for Source Term Modeling in Fluoride-salt-cooled High-temperature Reactors

The SAM code is under development and supported by DOE-NE’s Nuclear Energy Advanced Modeling and Simulation program as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. These advanced reactor concepts incorporate novel and improved approaches to achieve safety and economic feasibility. Many concepts leverage higher operating temperatures for improved efficiency with a variety of coolants and structural materials to support those needs. Such design choices may also pose unique radiological source term risks. There are continuous efforts to incorporate new physics and phenomena relevant to advanced reactor concepts, including capabilities to enable system-level source term modeling. The Pebble-Bed Fluoride-salt-cooled High-temperature Reactor (PB-FHR) is a promising candidate among advanced nuclear reactor concepts with its improved passive safety characteristics and high thermal efficiency. In addition to past efforts to support the development and utilization of SAM for PB-FHR safety analysis, the species transport modeling capabilities in SAM have been extended to simulate source term phenomena in concepts like the PB-FHR. The PB-FHR concept utilizes pebble-form TRISO fuel and FLiBe salt coolant to provide robust barriers to the release of almost all fission products and radiological source terms. However, tritium poses a unique radiological risk due to its significant production from neutron interactions with 6 Li and 9 Be in the FLiBe as well as its high mobility at elevated temperatures where it can permeate through structural metals. Secondly, the use of graphite at elevated temperatures poses the risk of oxidation damage to any structural components or fuel pebbles that are exposed to accidental air ingress. This report summarizes progress made in modeling these phenomena in SAM, which leverages the system-level multi-physics thermal hydraulic simulation to support an effective engineering-scale source term modeling capability. A tritium transport model is developed to simulate the various transport pathways in FHR and MSR concepts and is presented through verification and validation examples as well as demonstrations simulating experimental test loops and reference plant FHR models. A graphite oxidation model is also introduced to calculate local oxidation rates and is presented with initial validation results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Thermal properties measurement of TRISO particle coatings from room temperature to 900 °C using laser-based thermoreflectance methods

The thermal properties of tristructural isotropic (TRISO) particle coatings have been measured using laser-based thermoreflectance methods from room temperature to 900 °C. At room temperature the pyrocarbon coatings have comparable thermal conductivities below 10 W·m -1 ·K -1 , whereas the SiC coating has a thermal conductivity around 90 W·m -1 ·K -1 . The thermal diffusivities of all coatings display significant reduction with increasing temperature. Finally, the thermal conductivity of SiC decreases by more than 25% above 800 °C and the thermal conductivities of pyrocarbons increase moderately with the temperature, displaying similar changes to that of amorphous graphite at elevated temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Effect of microstructure and temperature on nuclear graphite oxidation using the 3D Random Pore Model

This work compared the 3D Random Pore Model (3D-RPM) with experimental and characterization data to systematically study the effect of nuclear graphite microstructure and air oxidation temperature. It is well known that oxidation-induced weight loss at elevated temperatures degrades graphite structure and properties; however, a fundamental understanding of the role of graphite microstructure is still unclear. In this work, three diverse grades of nuclear graphite—IG-110, NBG-18, and PCEA—were examined and tested at air oxidation temperatures of 600, 650, 700, and 750 °C following ASTM D7542. Here, the 3D-RPM reproduced the microstructure and temperature dependence of mass loss curves for these grades. For the first time, measurements and modeling were combined to show how bulk density and the amount and spatial distribution of open and closed porosity affects oxidation. IG-110 was less dense and had an open pore network that is finer and more uniformly spread, and showed fastest oxidation. PCEA porosity was less uniform and showed less oxidation, while NBG-18 was more dense, had the least fine and uniform porosity, and showed the slowest oxidation. In general, oxidation proceeds faster if open porosity is more uniformly distributed and can incrementally access closed pore surface area with more ease.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Optical and electronic properties of sub-surface conducting layers in diamond created by MeV B-implantation at elevated temperatures

Boron implantation with in-situ dynamic annealing is used to produce highly conductive sub-surface layers in type IIa (100) diamond plates for the search of a superconducting phase transition. Here, we demonstrate that high-fluence MeV ion-implantation, at elevated temperatures avoids graphitization and can be used to achieve doping densities of 6 at. %. In order to quantify the diamond crystal damage associated with implantation Raman spectroscopy was performed, demonstrating high temperature annealing recovers the lattice. Additionally, low-temperature electronic transport measurements show evidence of charge carrier densities close to the metal-insulator-transition. After electronic characterization, secondary ion mass spectrometry was performed to map out the ion profile of the implanted plates. The analysis shows close agreement with the simulated ion-profile assuming scaling factors that take into account an average change in diamond density due to device fabrication. Finally, the data show that boron diffusion is negligible during the high temperature annealing process.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2015

Enhancement of oxidation resistance of graphite foams by polymer derived-silicon carbide coating for concentrated solar power applications

Graphite foam with extremely high thermal conductivity has been investigated to enhance heat transfer of latent heat thermal energy storage (LHTES) systems. However, the use of graphite foam for elevated temperature applications (>600 °C) is limited due to poor oxidation resistance of graphite. In the present study, oxidation resistance of graphite foam coated with silicon carbide (SiC) was investigated. A pre-ceramic polymer derived coating (PDC) method was used to form a SiC coating on the graphite foams. Post coating deposition, the samples were analyzed by scanning electron microscopy and energy dispersive spectroscopy. The oxidation resistance of PDC-SiC coating was quantified by measuring the weight of the samples at several measuring points. The experiments were conducted under static argon atmosphere in a furnace. After the experiments, oxidation rates (%/hour) were calculated to predict the lifetime of the graphite foams. The experimental results showed that the PDC-SiC coating could prevent the oxidation of graphite foam under static argon atmosphere up to 900 °C.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Fabricating Fe nanocrystals via encapsulation at the graphite surface

In this paper, the authors describe the conditions under which Fe forms encapsulated nanocrystals beneath the surface of graphite, and they characterize these islands (graphite + Fe) thoroughly. The authors use the experimental techniques of scanning tunneling microscopy (STM) plus x-ray photoelectron spectroscopy (XPS) and the computational technique of density functional theory (DFT). Necessary conditions for encapsulation are preexisting ion-induced defects in the graphite substrate and elevated deposition temperature of 875–900 K. Evidence of encapsulation consists of atomically resolved STM images of a carbon lattice, both on top of the islands and on the sloping sides. The nature of the images indicates that this carbon lattice corresponds to a graphene blanket consisting of more than one graphene sheet that drapes continuously from the top of the island to the graphite substrate. The formation of iron carbide is not observed based on XPS. Shapes of the island footprints are consistent with metallic Fe, predominantly in the hcp or fcc form, though larger islands tend toward bcc. Island structures with hexagonally close-packed lateral hcp or fcc planes are stabilized by their excellent lattice match with the graphite substrate. Evolution of island density with prolonged deposition time provides evidence of coarsening, perhaps via Smoluchowski ripening. The encapsulated Fe clusters are stable in air at room temperature, protected by smaller Fe clusters that decorate defect sites and block permeation of gases. DFT shows that two configurations of Fe are more stable within the gallery than adsorbed on top of the surface: a single atom of Fe and a film (slab) of metallic Fe. Comparison with other metals shows that encapsulated Fe is similar to Cu but dissimilar to Ru or Dy, leading the authors to conclude that carbon dissolution in the metal does not play a role in encapsulation.

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 OSTI2020

Graphite Licensing (ASME) and Testing (ASTM) Technical Assistance

DOE ART Graphite R&D program is currently involved in developing the structural code for licensing the new High Temperature Reactor (HTR) designs through the ASME Boiler Pressure Vessel Code (BPVC). The basic structural design code for graphite core components has been written, approved, and within the BPVC since 2012. However, several areas within the graphite code remain to be improved, specifically the addition of material property changes resulting from environmental degradation (neutron flux and oxidation). The incorporation of this data, which can significantly affect the graphite components and alter the structural integrity of the core during operation, is proving to be an important activity. How this data will be incorporated within the code and how it will be used for license applications will be critical to a number of HTR applicants in the future. These modifications to the code must take into account the unique irradiation and oxidation response of the wide spectrum of available nuclear graphite grades currently under consideration for HTR applications. Additionally, the ASME BPVC requires material property testing within the expected operational conditions of the HTR core including neutron radiation flux, elevated temperatures, and after chronic/acute oxidation of the graphite components. Currently, no standardized graphite testing methods are approved for these environmental conditions. New American Society for Testing and Materials (ASTM) standardized tests for elevated temperature testing, small sample testing, and testing after oxidation must be developed to support the requirements within the ASME code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Investigation of the Electrode-Electrolyte Interphase in Ester-Based Electrolytes in NCM523/Graphite Cells

The use of methyl acetate (MA) and methyl propionate (MP) as co-solvents in carbonate/LiPF 6 electrolytes have been investigated in LiNi 0.5 Co 0.2 Mn 0.3 O 2 (NCM523)/graphite cells to improve electrochemical performance at low temperature (−20 °C). Consistent with previous investigations, improved low temperature performance was observed with all electrolyte formulations containing esters. Detailed ex situ surface analysis of the surface films generated with the ester-based electrolytes containing electrolyte additives reveal significant changes to the SEI on graphite upon cycling at moderately elevated temperature (45 °C). In the presence of vinylene carbonate (VC) lithium plating is observed upon cycling at low temperature (−20 °C) which leads to a thicker solid electrolyte interface (SEI) upon cycling at 45 °C. Incorporation of fluoroethylene carbonate (FEC) as an electrolyte additive results in the generation of a thin SEI and alters the structure of the surface film on the NCM523 cathode resulting in stable cycling over a wide temperature range.

Rodrigo, Nuwanthi D. (ORCID:0000000343129536)↗
DOE OSTI2026

Tribological behavior of nuclear graphite in high-temperature inert environment

This report formally documents the completion of the Advanced Reactor Technologies Level 2 Milestone (M2AT-26OR0605058), “Complete initial wear/abrasion studies,” due June 1, 2026. The work presented in this report expend upon our previously published manuscript titled Sliding friction and wear behavior of nuclear graphite in high temperature inert environment: Influence of contact load, speed and temperature. The sliding friction and wear of self-mated ET-10 nuclear graphite were characterized across a range of elevated temperatures (650°C and 750°C), contact loads (20 N and 40 N), and sliding speeds (1 mm/s and 10 mm/s) within a controlled argon environment. Tribological data combined with advanced morphological characterization were used to provide a detailed mechanistic framework for the frictional and wear behavior of nuclear graphite in high-temperature inert conditions. Furthermore, the report evaluates the inherent limitations of bench-scale characterization and addresses the critical disparities between laboratory findings and the complex, in-service friction and wear behavior of fuel pebbles within a reactor environment.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Sulfur-Doped Carbon Support Boosts CO2RR Activity of Ag Electrocatalysts

For presentation at the 70th AVS International Symposium and Exhibition. In this work, we show that the activity of Ag electrocatalysts for electrochemical CO2 to CO conversion is improved when supported on sulfur-doped (S-doped) carbon materials. S-doped carbon support was created by treating the heavily sputtered, highly oriented pyrolytic graphite (HOPG) in H2S at elevated temperatures, as confirmed by the S 2p X-ray photoelectron spectroscopy (XPS) peak. Scanning tunneling microscopy (STM) images indicated that Ag nanoparticles supported on S-doped HOPG had similar size distributions as those supported on sulfur-free (S-free) HOPG. While both catalysts reached > 90% CO Faradaic efficiency (FECO) at E = -1.3 V vs. the reversible hydrogen electrode (RHE) in the CO2 reduction reaction (CO2RR), Ag catalysts supported on S-doped HOPG demonstrated 70% higher CO turnover frequency (TOFCO = 3.4 CO/atomAg/s) than those supported on S-free HOPG (TOFCO = 2.0 CO/atomAg/s). Preliminary calculations based on density functional theory (DFT) indicated a more favorable energetic pathway of CO2-to-CO at the C-S-Ag interface, tentatively consistent with experiments. These results hint at a new approach to design active and selective electrocatalysts for CO2 conversion.

Deng, Xingyi↗
DOE OSTI2020

Low-temperature proton irradiation damage of isotropic nuclear grade IG-430 graphite

IG-430, a fine-grained, isotropic graphite grade is a promising candidate for the future Very High Temperature Reactors (VHTR). IG-430 which provides higher density, strength, and thermal conductivity, has already been developed as a graphite for next-generation HTGR, and is expected to be employed. This graphite grade, however, is lacking enough database that is needed for design. The present study aims to enhance the database with experimental data focusing on the low temperature regime (90–210 °C) by using 120–200 MeV protons to irradiate the IG-430 graphite to peak fluence of ~1.2 10 25 m –2 . It is anticipated that radiation-induced changes in the graphite properties and damage to be more pronounced in this low temperature regime than in elevated temperatures where damage annealing is taking place simultaneously. IG-430 graphite was characterized following irradiation for mechanical property changes (modulus and strength), dimensional stability and irradiation-induced growth as well as microstructural changes using high energy X-rays and different X-ray diffraction techniques. In assessing proton irradiation effects on the IG-430 graphite grade, comparison of radiation effects was made with the IG-43 grade, the un-purified version of IG-430, as well as other isotropic graphite grades. IG-430 was shown in this study to be better graphitized than other isotropic graphite grades. The study also revealed that during proton irradiation at low temperatures (~100 °C) the IG-430 exhibits stored energy release.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Effects of Synthesis Conditions on the Structure and Conductivity of Hydrogen-Substituted Graphdiyne

This study investigates how synthesis conditions influence the structure and conductivity of hydrogen-substituted graphdiyne (HsGDY). By varying the reaction temperature and solvent, we find that small changes in conditions markedly affect triple-bond retention and electronic continuity. Solid-state 13 C NMR and Raman spectroscopy reveal that elevated temperatures drive alkyne loss and partial graphitization, with N,N-dimethylformamide (DMF) promoting faster degradation than pyridine. The resulting decline in alkyne content directly correlates with reduced conductivity, indicating that preserving conjugation is essential for charge transport. These findings clarify how the synthetic environment governs the structural and electronic evolution of graphdiyne frameworks, providing insight into the controlled preparation of conjugated carbon networks.

alkyne retention↗
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↗