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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 OSTI2025

Summary of Graphite Data Stored within NDMAS

The Graphite Technology Development Project provides data to support the design of graphite core components within specific reactor service conditions of the next generation of high-temperature, gas-cooled nuclear reactors. Physical, mechanical, and thermal properties of nuclear grade graphite were characterized for specimens that were unirradiated, irradiated, and irradiated under various stress conditions. The material properties include diameter, length, mass, density, compressive strength, tensile strength, flexural strength, modulus, resistivity, thermal diffusivity, and thermal expansion coefficient. Baseline graphite specimens are unirradiated from different grades (2114, IG-110, NBG-17, NBG-18, and PCEA) and different types used in different characterization tests (compressive, flexural, tensile, one-inch cylinder, and quarter-inch cylinder). The Advanced Graphite Creep (AGC) irradiation specimens are from a much larger number of grades and cylinder types (creep, piggyback, and pencil). For baseline graphite, characterization data for 7,756 specimens extracted from thirteen graphite billets were captured to the NDMAS database. For AGC experiments, four irradiation campaigns have been completed: AGC-1, AGC-2, AGC-3, and AGC-4. The ongoing HDG-1 (High-Dose Graphite) experiment, which began irradiation with Cycle 168B on August 26, 2020, includes specimens previously irradiated in AGC-2 in addition to the specimens originally destined for AGC-5.. Besides the characterization data, the AGC data includes irradiation monitoring and physics data representing the irradiation conditions of AGC specimens. Currently, all data for AGC-1, AGC-2, and AGC-3 have been captured to the NDMAS database. Only AGC-4 pre-irradiation and irradiation monitoring data have been captured, and HDG-1 pre-irradiation data are in process of being captured for unirradiated specimens. To date, a total of 38,149 material property records have been captured into NDMAS database for baseline and AGC specimens. All characterization data are qualified for use according to their perspective data verification reports. For the AGC irradiation campaigns, a total of 41,502 qualified physics calculation records were added for AGC-1, AGC-2, and AGC-3. Finally, a total of 173,698,505 AGC irradiation monitoring data records (thermocouple temperature, gas flow rate, gas pressure, gas moisture, applied load, and specimen displacement) have been captured to NDMAS; the majority of those records (~94%) are qualified data records.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Sustainable Li-ion anode material from Fe-catalyzed graphitization of paper waste

We report a novel method for the conversion of paper towel waste to biographite anode material is developed and optimized for use in Li-ion batteries. The surge in demand for Li-ion battery anode materials coupled with the unsustainable and inefficient methods of producing battery-grade graphite necessitate alternative carbon feedstocks and graphitization technologies. Paper waste (PW) is identified as a suitable carbon feedstock for iron-catalyzed graphitization due to its sustainability, low cost, low ash content, and ample supply for the intended end use. A Box Behnken experimental design for statistical optimization is pursued for untreated and pre-carbonized PW with factors of temperature (1100-1300 °C), hold time (1-5 h), and iron catalyst loading (0.5-1.5x fixed carbon content) with biographite crystal size as the primary response variable. Temperature and iron catalyst loading are found to be significant factors, whereas hold time is found to be insignificant. Reversible capacities of the biographite anodes are found to be 340-355 mAh g-1 with 99% capacity retention over 100 cycles, indicating good electrochemical performance relative to commercial graphite anodes. The initial Coulombic efficiency of untreated and pre-carbonized biographites, however, are 77% and 75%, respectively, suggesting parasitic reactions including electrolyte decomposition.

25 ENERGY STORAGE↗
DOE OSTI2026

Are There Opportunities To Re-Think How We Manufacture Synthetic Graphite?

This manuscript contributes a Viewpoint article to ACS Sustainable Resource Management and discusses the graphite supply chain, growing mismatch between graphite demand and global manufacturing capacity, current graphite manufacturing technologies, and different feedstocks.

alternative carbon feedstocks↗
DOE OSTI2023

GRAPHITE–MOLTEN SALT CONSIDERATIONS FOR COMPONENTS IN NUCLEAR APPLICATIONS

The new High Temperature Reactor (HTR) designs being considered for future Gen IV nuclear reactor deployment include designs utilizing molten salt as the primary coolant. These molten-salt cooled, graphite core designs pose new material compatibility challenges that are not considered within the gas-cooled HTR designs that have been previously built and operated. While the Molten Salt Reactor Experiment (MSRE) demonstrated that the molten salt can be considered chemically inert to graphite the novel physical and thermal interactions that the molten salt poses may be just as impactful as the chemical reactivity. Specifically, molten salt intrusion into the open pore structure of nuclear graphite grades can provide additional internal stresses within the microstructure exacerbating the stress buildup from irradiation induced dimensional change. Additionally, designs using a molten salt containing liquid fuel could provide “hot spots” within graphite structural components causing local thermal stresses. Abrasion and erosion concerns are magnified with molten salt due to the extremely high density of the salts (some have higher densities than the structural graphite components). Finally, the graphite-graphite and fuel pebble-graphite tribological behavior are distinctly difference within the molten salt from the inert gas environments and must be investigated. These topics and others are currently under investigation within the DOE Advanced Reactor Technologies (ART) graphite program and will be discussed in depth.

Gallego, Nidia↗
DOE OSTI2022

GIF-VHTR Materials (Graphite) Project Plan: Progress since September 2021

¦1: Data, Design Methodology and Construction 1.1: Graphite Selection and Acquisition Strategy (Original tasks completed) 1.2: Graphite Properties (Virgin Materials) 1.3: Graphite Fracture Behaviour(Original tasks completed) 1.4: Graphite Oxidation Behaviour(Original tasks completed) 1.5: Graphite Component Testing 1.6: Graphite Irradiation Effects 1.7: Graphite Irradiation Induced Creep 1.8: Graphite Codes and Standards Development (ASME & ASTM) 1.9: Graphite BehaviourModels Development 1.10: Links to Existing Graphite Irradiation BehaviourDatabases 1.11: Materials Database ¦2: Operation and Inspection 2.1: Methodologies for determining requirements for inspection of graphite cores 2.2: Methodologies for ISI of graphite cores ¦3: Decommissioning and Disposal (New task added 2021) 3.1: Technical data for decommissioning and disposal issues

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Current and Future Technological Issues Challenges for Nuclear Graphite Components

Historical and current data requirements for component qualification, as-manufactured graphite material properties, Irradiated & degraded material issues, code rules – Construction & Operation, status of current ASME code rules, progress in design rules, degradation, construction vs. operation, what should we be planning, and new technical areas getting started.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

HDG-1 Experiment Irradiation Monitoring Data Qualification Final Report

SUMMARY The U.S. Department of Energy (DOE) Advanced Reactor Technologies (ART) Graphite Research and Development (GRD) Program is conducting a series of six experiments to quantify the effects of irradiation on nuclear-grade graphite. This report documents the qualification of irradiation monitoring data for the fifth experiment, High Dose Graphite-1 (HDG-1). Qualified monitoring data are required by the ART program to support the design and licensing of the first high-temperature reactor (HTR) nuclear plant. Data are classified as Qualified if they meet the usage requirements described in the experiment planning and quality assurance (QA) documents, Failed if they do not meet those requirements and provide no usable information, or Trend if they do not fully meet all requirements but still provide useful information subject to an assessment of how any deficiencies may affect a particular use of the data. HDG-1 irradiation began with Advanced Test Reactor (ATR) Cycle 168B on August 24, 2020, and concluded after Cycle 173C on January 27, 2025. The HDG-1 capsule was removed from the reactor core twice—during core internal change (CIC) Cycle 170A and powered axial locator mechanism (PALM) Cycle 172A—to prevent overheating of the graphite specimens during high-power PALM cycles. The capsule was therefore irradiated during a total of seven normal ATR cycles: 168B, 169A, 171A, 171B, 173A, 173B, and 173C. Irradiation monitoring data evaluated in this report include thermocouple (TC) temperature, gas flow rate, gas moisture, gas pressure, specimen load, and graphite stack displacement. Temperature. A total of 14,508,065 TC temperature records were captured. Of these, 13,901,785 (95.8%) are Qualified and 606,280 (4.2%) are Failed. The principal source of failed temperature data was the instrument failure of TC-9 (Zone 2) on June 24, 2024, and TC-10 (Zone 1) on July 5, 2024, near the end of Cycle 173A, which resulted in 595,554 Failed readings. An additional 379 missing values and 10,347 slightly negative values from TC-13 during ATR outages are also Failed. Neither TC-9 nor TC-10 was used as a temperature-control TC, and their failures did not compromise capsule condition monitoring. Correlation analysis of all 13 TCs found no evidence of virtual junction formation. Control chart analysis revealed clear downward drift of approximately 80°C for TC-6 (Zone 3) relative to other stable TCs, and possible downward drift of approximately 60°C for TC-13 relative to the Zone 5 control TC (TC-1), though TC-13 remained consistent with the Zone 2 control TC (TC-12). Gas flow. A total of 20,088,090 gas flow rate records were captured. Of these, 19,941,463 (99.3%) are Qualified and 146,627 (0.7%) are Failed due to missing values. All argon, helium, and total gas flow data were within expected ranges throughout the irradiation. Gas moisture. A total of 1,116,005 outlet gas moisture values were captured. Of these, 1,101,421 (98.7%) are Qualified and 14,584 (1.3%) are Failed, comprising 14,556 out-of-range values and 28 missing values. The out-of-range moisture values exceeded 22,000 ppmv for approximately 1 week at the beginning of Cycle 173A, when accumulated moisture evaporated after the capsule was retrieved from water storage during PALM Cycle 172A and reinserted into the east flux trap. Moisture levels returned to below 25 ppmv for the remaining three cycles, and the transient high-moisture event did not affect the integrity of specimen irradiation. Gas pressure. A total of 7,812,035 gas pressure values were captured. Of these, 6,642,048 (85.0%) are Qualified and 1,169,987 (15.0%) outlet pressure values are Failed, comprising 718,537 zero outlet pressure values due to sensor failure from Cycle 168B through Cycle 171B, 54,550 missing values, and 396,900 too-low outlet pressure values, ranging from 1.1 to 1.6 psia after sensor replacement during Cycle 173A. Load. A total of 6,696,030 load values were captured. Of these, 6,694,580 (99.98%) are Qualified and 1,450 (0.02%) are Failed due to missing values. Applied loads to the six specimen stacks were stable throughout the irradiation. Stack displacement. A total of 6,696,030 displacement values were captured. Of these, 5,713,297 (85.32%) are Qualified and 3,781 (0.06%) are Failed due to missing values. Stack displacement increased consistently throughout the irradiation, reaching approximately 3.08 in. for Channels 5 and 6 by the end of irradiation. 978,952 (14.62%) substantially elevated displacements observed for Channel 6 beginning in Cycle 171A and for Channel 5 beginning in Cycle 173A are assigned Trend status. Raising pressure. A total of 1,115,999 raising pressure values were captured. Of these, 1,115,430 (99.95%) are Qualified and 569 (0.05%) are Failed due to missing values. Ram pressure. A total of 6,696,030 ram pressure values were captured. Of these, 6,692,249 (99.95%) are Qualified and 3,484 (0.05%) are Failed due to missing values. Stack raising was perf

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

TRISO Particle Isolation and Graphite Removal Using SRNL Vapor Digestion Technology

Tri-structural isotropic fueled reactors are planned to come online in the next decade, but there is not currently a widely agreed upon disposition pathway for this new fuel stream. For long term used nuclear fuel storage, there would be a significant benefit if the waste volume could be reduced or mitigated. Most of the volume from used TRISO fuel contains graphite. SRNL currently has an active patent which describes one pathway to digest the graphite. One benefit in the utilization of this pathway is that multiple cylindrical pieces can be stacked end-to-end in a reaction tube, then nitric acid and water vapor can be flowed through the tube, and the weight change of the individual cores can provide both a reaction profile and overall oxidant use efficiency. The reaction of nitric acid and graphite passes through a series of intermediate products – NO 2 , NO, and N 2 O – from reaction and decomposition to eventually form N2. The data shows that NO2 grows in with increasing temperature between 500-600°C and then decreases by way of either reaction with graphite or decomposition. Nitric oxide, a reaction and thermal decomposition product of NO 2 , exhibits a consistent decline as a function of temperature, which is consistent with the literature. Similarly, the concentrations of N 2 O and N 2 increase as a function of temperature as their reactions with graphite become more favorable.

Advanced Reactors↗
DOE OSTI2023

A novel method for quantifying irradiation damage in nuclear graphite using Raman spectroscopy

Raman spectroscopy has long been used in studying irradiation damage in carbon/graphite materials. It is however unclear if the measurements from different types of materials are directly comparable. Further, decoupling the contribution from irradiation temperature and fluence to the total damage in nuclear graphite possessing irradiation damage gradient is currently not readily feasible as it requires a complete set of test reactor irradiation experiments over the relevant temperature and fluence range. A novel methodology has therefore been proposed and developed to quantify total irradiation damage evolution at crystal level based on graphite Raman G-band position shift. Specifically, G-band positions derived from a large number of spectra collected from the fractured surface of microfine-grained POCO ZXF-5Q graphite possessing proton irradiation damage gradient were plotted with open literature Raman data on HOPG, BEPO (AXGP graphite), PCEA and IG-110 graphite as a function of dpa. The total damage level within 2σ beam radius in this POCO graphite was estimated to be equivalent to ~2–5 dpa at ~350–370 °C. Derived G-band positions were then mapped to the three-stage amorphization trajectory model indicating beam centre area has entered the second stage, i.e., transitioning from nanocrystalline graphite into amorphous carbon. G-band position relative shift (ΔG) curve with a ‘turn-around’ peak as a function of total damage can be used for in-service lifetime prediction. Importantly, the developed methodology has the potential to ‘unify’ total damage levels across different grades of nuclear graphite caused by ion, neutron and proton irradiation at different temperatures.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Graphite for Advanced Nuclear Reactors: Deployment Readiness Review

Historically, graphite has been used in numerous reactor technologies, including research/test reactors and, commercially, in advanced gas reactors. These non-metallic materials can play a key role as internal core structures, reflectors, or neutron moderators, making them important for deployment of certain advanced reactor technologies. This report explores the industry readiness for graphite material deployment, documenting Codes and Standards applicable to their design, qualification, and manufacturing. Discussions also examine the manufacturing processes, aging/degradation, inspection techniques, and disposal options. The report covers the current status, documents gaps, and proposes some conclusions about approaches to managing some of the current gaps.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Upgrading of Raw Coal and Coal Waste for Coal-Derived Graphene Process

Conference presentation at 47th International Technical Conference on Clean Energy (Clearwater Clean Energy Conference), Clearwater, Florida, July 23–27, 2023. The University of North Dakota Energy & Environmental Research Center (EERC) conducted a laboratory-scale coal-derived graphene (CDG) project focused on developing a technological process for making graphite from four U.S. domestic coals and coal wastes. Coal and coal waste preprocessing methods were developed and applied to clean and upgrade the coal precursors prior to graphitization and subsequent conversion to graphene products. Carbonization and graphitization of these preprocessed coals and coal wastes has produced graphite, which was used to make graphene oxide (GO) and reduced graphene oxide (rGO). Graphene quantum dots (GQDs) were also made from the raw and upgraded coal precursors.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2025

Development of the technological process for the IGR reactor's highly-enriched irradiated uranium-graphite fuel immobilization

The immobilization of irradiated highly enriched uranium (HEU) fuel is a critical component of nuclear waste management and non-proliferation efforts. In Kazakhstan, at National Nuclear Center of the Republic of Kazakhstan special attention is given to managing legacy HEU fuel from research reactors. One such case involves the IGR research reactor, whose first core containing irradiated HEU uranium-graphite fuel was operated from 1961 to 1966 and removed following reactor modernization. This fuel now requires a reliable and secure immobilization strategy. Here, this paper presents the development of a technological process for immobilizing this fuel to reduce its enrichment to below 20 % in terms of 235U content. The proposed method involves down-blending irradiated HEU fuel with depleted uranium, followed by encapsulation in a Portland cement matrix. Full-scale experiments were conducted to assess the uniformity of uranium distribution within the matrix. The results confirm the effectiveness of this approach, ensuring reliable immobilization of fuel in accordance with international requirements, including IAEA standards and Kazakhstan's regulatory framework. These findings contribute to the broader effort of adapting immobilization strategies for the safe management of spent fuel from research reactors.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2026

Status on Microstructural Studies of Nuclear Graphite

This report documents the completion of the Advanced Reactor Technologies (ART) Level 3 Milestone (M3AT-26OR0605054), “Provide status on microstructural studies of nuclear graphite,” due May 1, 2026. The contents of this report summarize recent progress in the development of the Nuclear Graphite Microstructure Library, including datasets that have been generated, curated, and submitted for publication, as part of the broader effort to establish and integrate the library within the NDMAS platform.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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 OSTI2024

Carbon Ore-Derived Critical Materials for Clean Energy Technologies

Conference presentation at American Institute of Chemical Engineers (AIChE) Annual Meeting, San Diego, California, October 27–31, 2024. Trends in the manufacture of electric vehicles that use graphite-based LIBs are rising steeply in the United States and globally, but the U.S. domestic supply chains for graphite, the largest component in an LIB by mass, is severely limited and faces complicated geopolitical dynamics with foreign sources. Consequently, the United States has designated graphite as a critical mineral to focus attention and resources to develop technologies to meet the challenge of limited domestic graphite supply chains. Results obtained so far based on the UCOP process have successfully validated the technology at the laboratory scale, with the produced graphite material showing up to 95% degree of graphitization, high carbon purity of ~99.98%, residual ash content of ≤0.02%, negligible moisture, low trace elements, and high electrochemical stability. These results suggest that the emerging UCOP technology is a promising approach to effectively synthesize high-quality graphite from abundant coal and coal waste resources in the United States to create a sustainable domestic critical graphite supply chain. A brief description of the status of UCOP process development and representative results will be presented.

01 COAL, LIGNITE, AND PEAT↗