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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 19 records
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 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 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 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 OSTI2026

3D TRISO particle-explicit compact meshing

The TRI-structural ISOtropic (TRISO) layered fuel particle is a robust nuclear fuel form offering enhanced safety and performance for advanced reactor concepts, including high-temperature gas-cooled reactors and other Generation IV designs. These poppy-seed-sized particles are embedded in a graphite matrix to form fuel elements that must withstand elevated temperatures and high burn-up levels. The heterogeneous nature of these fuel elements — comprising thousands of randomly distributed TRISO particles — produces complex stress fields and thermal gradients that one- and two-dimensional models cannot accurately capture. While three-dimensional modeling has improved predictions of dimensional changes, internal pressure buildup, and fission product transport under irradiation, current approaches rely on homogenized material properties that are known to have considerable divergence from experimental observations. This work presents a methodology for optimized random packing of TRISO fuel compacts and full three-dimensional mesh generation within the BISON fuel performance code, with each particle coating layer individually discretized. The resulting mesh was demonstrated through heat conduction simulations under representative in-reactor operating conditions, showing strong agreement with expected behavior. This capability enables detailed analysis of particle-to-particle interactions, matrix cracking mechanisms, and the statistical distribution of coating layer failures — all of which directly govern fuel performance and safety margins.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Microwave-Assisted Reactive CO 2 Capture with the SrCO 3 -Graphite System

This study details the initial development of a microwave heat-driven reactive CO 2 capture (RCC) process using the SrO/SrCO 3 cycle and graphitic carbon to absorb ppm-levels of CO 2 from humidified room temperature air and selectively convert it into CO. By combining first-principles density functional theory (DFT) simulations with thermogravimetric analysis (TGA) and X-ray diffraction (XRD) verification, it is demonstrated that SrO spontaneously absorbs moisture to generate Sr(OH) 2 followed by Sr(OH) 2 *1H 2 O, which then can spontaneously react with atmospheric levels of CO 2 to form SrCO 3 . The resulting SrCO 3 can then react with solid carbon at elevated temperatures to selectively produce CO and regenerate SrO. Graphitic carbon is an excellent microwave absorber that can quickly generate temperatures approaching 1000 °C and then cool to room temperature within minutes, potentially allowing for integration with intermittent electricity. It is shown that using microwaves to selectively heat a mixture of graphitic carbon and SrCO 3 produced CO with 85 ± 3% selectivity and stable performance over ten cycles. The rapid heating for release and room temperature CO 2 uptake processes appeared to prevent performance loss from particle sintering typically observed with traditional thermal systems. In conclusion, these results demonstrate a new RCC approach that adds to the growing number of potential technologies available for converting CO 2 to useful chemicals.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Low-Temperature Production of Battery Grade Graphite from Coal with Recovery and Reuse of the Catalyst

Synthetic graphite is made predominantly from a petroleum-derived needle coke using the Acheson method. This involves heating the needle coke at elevated temperatures of ~ 3000 °C for more than 7 days. High-temperature heating is the most technically challenging, expensive, and energy-intensive aspect of graphite manufacturing. We will present a strategy for synthesizing high-quality battery-grade graphite powder from coal using a low-temperature catalytic graphitization process. The use of a catalyst drops the processing temperatures to 1500 °C and processing times to a few hours. The graphite produced with this process has been tested as a battery anode and has been shown to outperform anodes made with commercially sourced graphite materials. We will also present a sustainable method to recover and reuse the catalyst and acid used to retrieve the catalyst. This overcomes a long-standing technical hurdle associated with using catalysts for manufacturing graphite.

Kim, Ki-Joong↗
DOE OSTI2026

Demonstration of an All-Refractory Corrosion-Resistant Molten Salt Pumping and Storage Infrastructure Up to 950 degrees C

Concentrating solar power (CSP) with thermal energy storage (TES) has an estimated cost similar to solar photovoltaics (PV) with lithium-ion batteries (LIB), but CSP + TES has the potential for significantly reduced cost by operating above 700 degrees C. However, this requires a TES medium and containment infrastructure which are chemically compatible and do not degrade above 700 degrees C. MgCl2-KCl-NaCl (MKN) salt is a promising medium, but when MKN salt has small amounts of water and oxygen dissolved in it, it excessively corrodes conventional commercial alloys. Here, we consider refractory materials including graphite, carbon-carbon composite (C/C), and molybdenum because they maintain high mechanical strength at elevated temperatures and are expected to resist corrosion by MKN salt. While selection of refractory materials is often constrained by the need to remain chemically stable in air, this work uses an enclosure filled with inert gas to allow the use of materials like graphite and molybdenum. Notably, graphite components can seal against liquids unlike many brittle refractory materials. We demonstrate a centrifugal pump and mechanical seals made of refractory materials in a laboratory-scale circulation loop, which successfully operated continuously for more than 49 h at temperatures ranging from 750 degrees C to 950 degrees C with no mechanical failures or chemical degradation of the refractory materials. To the best of our knowledge, this is the highest temperature molten salt circulation loop that has been successfully developed, and this refractory infrastructure allows for operation at even much higher temperatures (<1600 degrees C) if a storage medium with a lower vapor pressure is used. The principles of this architecture are relevant to other applications with high temperature flowing metals and salts including Generation IV nuclear fission reactors for power production and molten salt electrolysis reactors for metallurgical processing.

14 SOLAR ENERGY↗
DOE OSTI2025

Highly Crystalline Graphite Synthesis from Coal with a Sustainable Process

Synthetic graphite is made predominantly from a petroleum-derived needle coke using the Acheson method. This involves heating the needle coke at elevated temperatures of ~ 3000 °C for more than 7 days. High-temperature heating is the most technically challenging, expensive, and energy-intensive aspect of graphite manufacturing. We will present a strategy for synthesizing high-quality graphite powder from coal using a low-temperature catalytic graphitization process. The use of a catalyst drops the processing temperatures to 1500 °C and processing times to a few hours. The graphite produced with this process has been characterized and has been shown to have comparable physical/chemical properties with commercially sourced graphite materials. We will also present a sustainable method to recover and reuse the catalyst and acid used to retrieve the catalyst. This overcomes a long-standing technical hurdle associated with using catalysts for manufacturing graphite.

crystalline graphite powder↗
DOE OSTI2025

A high-temperature Rutherford Backscattering Spectrometry apparatus for in situ material characterization

A new methodology for high-temperature Rutherford Backscattering Spectrometry (HT-RBS) has been developed to enable in situ material characterization at elevated temperatures. A 3.5 MeV proton beam penetrates a 10-µm-thick 316L stainless steel foil mounted on a graphite substrate, with backscattered signals detected using an HT-RBS system. Conventional semiconductor detectors, primarily based on silicon, suffer significant performance degradation at temperatures higher than ~ 60 °C due to increased leakage current and noise, leading to signal distortion and failure. Here, to preserve spectral quality, a 5 µm aluminum foil shields the detector from thermal radiation, allowing reliable operation up to 900 °C at the target. A rotatable shutter provides additional thermal isolation during data collection pauses. In situ measurements of areal density changes of 316L stainless steel were conducted to validate the technique, revealing consistency with the known thermal expansion coefficient. The method facilitates seamless switching between irradiation and analysis, enabling continuous studies. This approach supports in situ investigations of diffusion, void swelling, creep, and corrosion, offering a versatile tool for advanced materials research.

36 MATERIALS SCIENCE↗
DOE OSTI2024

On the product phases and the reaction kinetics of carbothermic reduction of UO 2 +C at relatively low temperatures

The synthesis of UC using carbothermic reduction of UO 2 and C mixtures has been well studied at high temperatures. However, the product phase behavior of carbothermic reduction at low temperatures (≤1773 K) is not well studied. Such a study is important as low temperatures permit single phase UC synthesis without forming secondary higher carbides, and it further supports the knowledge base of the process that needs to be used for transuranic elements such as plutonium that have high vapor pressures at elevated temperatures. Therefore, a low temperature carbothermic reduction of two different C/UO 2 molar ratios under inert and reducing environments have been studied here. Two different sample holding crucibles, alumina (Al 2 O 3 ) and graphite, were also used here to differentiate the hypostoichiometric (UC 1-a ) and oxygen dissolved (UC 1-x O x ) uranium monocarbide phases adding more details on the two systems. Also, the reaction kinetics involved in the formation of UC via the carbothermic reduction of UO 2 +C using product phases instead of evolved gases such as carbon monoxide is reported here. Under inert atmospheres but with significant oxygen partial pressures, the low temperature carbothermic reduction of UO 2 +C produced up to 90 wt.% UC 1-x O x type oxycarbides as was confirmed by Xray powder diffraction. Reducing Ar-4%H 2 environments at these temperatures were not successful in synthesizing UC as it reduces the amount of C required for the carbothermic reduction, leaving UC phase at a non-equilibrium state. Inert atmospheres with low or negligible oxygen partial pressures on the other hand produced near stoichiometric UC at high phase purity, especially at 1673 – 1773 K temperature range. An activation energy of 377±75 kJmol -1 was also calculated using product phase concentrations of the carbothermic reduction of UO 2 +C under these inert Ar (g) atmospheres.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Deactivation of Mo/H-ZSM-5 in Microwave-Assisted and Thermal-Driven Methane Dehydroaromatization

A better understanding of catalyst deactivation is needed to improve catalyst design and performance in microwave-enhanced methane dehydroaromatization (MDA). Here, this study investigates the deactivation of a molybdenum supported H-ZSM-5 zeolite (Mo/H-ZSM-5) catalyst in MDA under microwave-heated conditions, comparing its performance to that of the same catalyst under conventional heating. While the microwave-assisted (MW) process achieved higher benzene yields, the catalyst experienced faster deactivation due to the selective and rapid deposition of coke within the pores of the zeolite, as confirmed through Brunauer–Emmett–Teller (BET), X-ray diffraction (XRD), ammonia-temperature programmed desorption (NH 3 -TPD), thermal gravimetric analysis (TGA), temperature programmed oxidation (TPO), and X-ray photoelectron spectroscopy (XPS) analyses. The quantification of total coke content via TGA/TPO and surface carbon (XPS) revealed that nearly twice as much coke was deposited on the catalyst under MW conditions compared to that on the conventionally heated material, and the coke exhibited a more conductive and graphitic nature. The accelerated deactivation rates were attributed to the formation of hot spots in the MW system, leading to enhanced coupling with coke formed in situ during the reaction and resulting in increased Mo reduction. Observations indicated that the CO activation used to carburize the catalyst prior to the reaction is not advantageous in the MW heating environment. The presence of large amounts of Mo oxides at elevated temperatures (through hot spots) exposed to methane leads to instability under the reaction conditions. Optimizing the activation environment and improvement of the Mo dispersion within the pores are potential strategies to improve catalyst stability.

Mo/H-ZSM-5 zeolite↗
DOE OSTI2025

Use of Constrained Gamma Emission Computed Tomography to Evaluate Fission Product Distributions in High-Temperature Materials from a TRISO Fuel Irradiation

An image reconstruction technique was developed to overcome problems with earlier methods of gamma emission computed tomography of graphite rings surrounding the AGR-3/4 TRISO fission product transport experiment. The profiles obtained from the tomography are compared with sampling done via radially resolved destructive sampling techniques. Generally, there is good agreement between profiles measured via destructive sampling of the rings and the tomographic profiles, though at low signal strengths, the tomographic profiles appear elevated.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Rational Electrolyte Design for Elevated-Temperature and Thermally Stable Lithium-Ion Batteries with Nickel-Rich Cathodes

As the energy density of lithium-ion batteries (LIBs) increases, the shortened cycle life and the increased safety hazard of LIBs are drawing increasing concerns. To address such challenges, a series of localized high-concentration electrolytes (LHCEs) based on a solvating-solvent mixture of tetramethylene sulfone and trimethyl phosphate and a high flash-point diluent 1H,1H,5H-octafluoropentyl 1,1,2,2-tetrafluoroethyl ether were designed. The LHCEs exhibited non-flammability and greatly suppressed heat release at high temperatures, which would potentially improve the safety performance of the LIBs. Moreover, the optimal LHCE achieved capacity retentions of 87.1% and 81.7% in graphite||LiNi 0.8 Mn 0.1 Co 0.1 O 2 cells after 500 cycles at 25 °C and 45 °C, respectively, which were significantly better than the conventional electrolyte, whose capacity retentions were only 75.2% and 38.5% under the same condition. Mechanistic studies revealed that the LHCE not only formed a more robust solid electrolyte interphase, but also exhibited improved anodic stability, compared with the conventional electrolyte. Further, this work sheds light in rational electrolyte design for high energy density LIBs with high battery performance and low safety concerns.

25 ENERGY STORAGE↗
DOE OSTI2025

Fuel performance analysis of fully-resolved TRISO compact

The TRi-structural ISOtropic (TRISO) fuel multilayered coating structure offers multiple barriers to fission product release, enhancing safety and performance. The heterogeneous nature of TRISO fuel compacts, comprising thousands of randomly distributed coated fuel particles embedded in a graphite matrix, creates intricate stress fields and thermal gradients that cannot be accurately modeled using simplified one-dimensional or homogenized approaches. Consequently, three-dimensional modeling enables the prediction of fuel compact dimensional changes, internal pressure buildup, and fission product transport pathways under diverse irradiation and thermal conditions. This capability facilitates detailed analysis of particle-to-particle interactions, matrix cracking mechanisms, and the statistical distribution of coating failures, which directly impact fuel performance and safety margins. This capability is particularly critical for advanced reactors, such as high-temperature gas-cooled reactors and other Generation IV reactor designs where TRISO fuel operates at elevated temperatures and burn-up levels. This work introduces a novel method to generate an optimized packing of TRISO compacts and a complete 3D mesh with random distribution of TRISO particles, which are discretized into each coating component layer.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

AGC-2 Graphite Preirradiation Data Analysis Report

This report describes the specimen loading order and documents all preirradiation examination material property measurement data for graphite specimens contained within the Second Advanced Graphite Capsule (AGC 2) irradiation capsule. The AGC 2 capsule is the second in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens in order to garner quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear grade graphites. This testing will ascertain the in service behavior of the graphite for pebble bed and prismatic very high temperature reactor designs. Similar to the First Advanced Graphite Capsule (AGC 1) preirradiation examination report, material property tests were conducted on specimens from 18 nuclear grade graphite types. However, AGC 2 tested an increased number of specimens (i.e., 512) prior to loading them into the AGC 2 irradiation assembly. All AGC 2 specimen testing was conducted at Idaho National Laboratory from July 2009 to August 2010. This report also details the specimen loading methodology for graphite specimens inside the AGC 2 irradiation capsule. The AGC 2 capsule design requires “matched pair” creep specimens that have similar dose levels above and below the neutron flux profile mid plane. This provides similar specimens with and without an applied load. Analysis in this document utilizes the neutron flux profile calculated for the AGC 2 capsule design, the capsule dimensions, and the size (i.e., length) of the selected graphite specimens to create a stacking order that produces “matched pairs” of graphite specimens above and below the AGC 2 capsule elevation mid point, thus providing specimens with similar neutron dose levels.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗