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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 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Interfacial electron-phonon coupling and quantum confinement in ultrathin Yb films on graphite

Interfacial electron-phonon coupling in ultrathin films has attracted much interest recently. Here, by combining angle-resolved photoemission spectroscopy and scanning tunneling microscopy, we report quantized electronic states and strong interfacial electron-phonon coupling in ultrathin Yb films on graphite. We observed clear kinks in the energy-momentum dispersion of quantum well states, and the kink positions agree well with the energies of optical phonons of graphite. The extracted coupling strength λ is largest for the thinnest film with a preferred (“magic”) thickness of four monolayers and exhibits a strong band dependence, which can be qualitatively accounted for by a simple model. The interfacial electron-phonon coupling also gives rise to characteristic steplike structures in the $dI/dV$ spectra, implying dominant coupling with the phonons with zero in-plane momentum. A Lifshitz transition occurs at higher coverage, where quantum well states derived mainly from 5d electrons dominate near the Fermi level and possess large effective mass (up to ~ 19 m e ). Here our results highlight the potentially important role of interfacial electron-phonon interaction for ultrathin films and provide spectroscopic insight to understand this cross-interface fermion-boson interaction.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Investigation of nonlinear mechanical behavior of two superfine-grained graphites with DIC assisted disc splitting test

Using standardized uniaxial tensile test specimens is not practical due to the limited volumes in irradiation capsules, molten salt degradation facilities, or oxidation apparatus. The ASTM International Standard Test Method for Tensile Strength Estimate by Disc Compression of Manufactured Graphite (ASTM D8289) was developed to provide a convenient way for estimating tensile strength. Unlike the traditional uniaxial tensile test in ASTM C749 (International Standard Test Method for Tensile Stress–Strain of Carbon and Graphite), which uses dog-bone shaped specimens larger than 12.95 mm × 120.65 mm, the ASTM D8289 standard uses smaller discs with diameters of 6–12.7 mm. A digital image correlation (DIC) system, which uses a full-field noncontact surface displacement measurement technique, was applied along with the ASTM D8289 disc splitting test on IG-110 samples and compared with previous measurements from Mersen 2114 samples. Results confirmed that the DIC technique can measure the surface displacement/strain on these small (Ø6 mm × 3 mm) graphite specimens with good repeatability. However, Mersen 2114 and IG-110 samples exhibited strain discrepancies when DIC measurements were compared with analytical and finite element simulation values. The loading history and strain results also indicated different mechanical behaviors between Mersen 2114 and IG-110, particularly the nonlinear behavior of the IG-110 samples. Good agreement was observed by comparing the splitting tensile strengths of two superfine-grained grades with results from other work. The specimen size effect is discussed when comparing the splitting tensile strength with corresponding uniaxial tensile strength of these two graphite grades.

Lin, Lianshan [ORNL] (ORCID:0000000203399219)↗
DOE OSTI2023

Baseline Graphite Characterization

Presentation will include baseline graphite characterization information, such as purpose and results, baseline property measurements, baseline graphite characterization method, baseline progress, split disc tensile strength: ASTM Tensile Strength Alternative, Identification of proper split disk fracture, baseline graphite ASTM D8289, horizontal orientation, digital image correlation, and DIC Analyzation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Evaluating design safety margins in the American Society of Mechanical Engineers graphite core components design-by-analysis assessments

Graphite is an important material being used for core components in next-generation high-temperature gas-cooled nuclear reactors. The selection of graphite grade for a specific Designer is a complex task, dependent on reactor conditions, component functionality, and required reliability. The American Society of Mechanical Engineers (ASME) provides two semi-probabilistic design-by-analysis assessments to evaluate graphite core components against design reliability targets. The simplified assessment uses a 2-parameter Weibull distribution to describe the graphite grade’s tensile-strength distribution to establish component stress limits. The full assessment uses the 3-parameter Weibull distribution and a modified Weakest-Link Theory approach to calculate a component design probability of failure. The paper defines recommended assessment rules, which are the as-written simplified assessment and the full assessment with parameter lower bounds, the modulus update with threshold reduction, and the 2027 grouping rules. Code rules are applied to three grades: 2114, IG-110, and NBG-18. The baseline margin calculation is developed using the experimental tensile dogbone specimen. Percent margin is defined as the percent reduction in the median experimental load to obtain the allowable load per ASME assessments. Under the recommended rules, the SRC–1 margin in the simplified assessment ranged from 40.2 % to 52.7 % among the grades in this study and from 36.1 % to 49.8 % in the full assessment. The full assessment only decreases the margins by 2.5–4.5 % for the SRC-1 components and 0–1.5 % for the SRC-2 components for this baseline case. Margin is inversely related to material median strength (i.e., the strongest grade, 2114, has the lowest margin).

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

UNDERSTANDING THE SEMI-PROBABILISTIC APPROACHES IN STRUCTURAL RELIABILITY USED TO SET DESIGN RELIABILITY TARGETS FOR GRAPHITE COMPONENTS USING ASME BPVC METHODS

Graphite is a quasi-brittle material, resulting in random variability in tensile strength distributions. To account for the random variability in strength, HHA-3000 of the ASME BPVC provides two semi-probabilistic methods for qualifying nuclear graphite components in the design stage, the simplified and full assessments. The full and simplified assessments apply statistical methods to engineering-based design problems. This is often referred to as reliability-based design. Reliability-based design (RBD) is a method to develop reliable designs by accounting for uncertainties and result in small chances of failure when also considering safety factors. RBDs provide reliability targets using semi-probabilistic approaches. RBD is implemented in ASME BPVC HHA-3000 for nuclear graphite components, but is not specific to that application. There has been much confusion around the methods implemented in ASME BPVC HHA-3000 for qualifying nuclear graphite components. To address the confusion, this paper takes a hierarchical approach. First, the general RBD framework is presented. Then, the semi-probabilistic methods and the underlying assumptions implemented in the assessments are presented. The semi-probabilistic methods are separated from the engineering modifications that have been made to the assessments. After building the framework and underlying assumptions, the specific methods in the full and simplified assessments are explained in three steps: inputs, methods, outputs. The methods are applied to an H-451 reflector block. Tensile strength properties for other graphite grades are provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Li-Ion Battery Electrode Contact Resistance Estimation by Mechanical Peel Test

Li-ion battery electrode electronic properties, including bulk conductivity and contact resistance, are critical parameters affecting cell performance and fast-charge capability. Contact resistance between the coating and current collector is often the largest electronic resistance in an electrode and is affected by chemical, microstructural, and interfacial variations. Direct measurements of contact resistance and bulk conductivity have proven to be challenging. In their absence, a mechanical electrode peel test is often used to compare adhesion and electrical contact resistance. However, using a micro-flexible-surface probe, contact resistance can be directly determined. Here, this work compares contact resistance and mechanical peel strength of multiple commercial-grade HE5050 and NCM523 cathodes and graphite and silicon anodes. It was found that peel strength correlates well with contact resistance in a carefully curated data set (p < 0.05) and in some situations may be a good metric to estimate electrical properties. However, there were distinct outliers in the data set, indicating that peel strength may not accurately reflect electrical properties when there is significant variation in electrode composition. These results illustrate the value of the micro-flexible-surface probe in quantifying contact resistance and bulk conductivity to better understand how battery composition and processing steps affect microstructure and resulting cell performance.

25 ENERGY STORAGE↗
DOE OSTI2021

Developing a New Polyolefin Precursor for Low-Cost, High-Strength Carbon Fiber

Current high-strength carbon fibers (CFs) are mostly derived from PAN polymer precursor, which are very expensive with limited usage in the high-end products. On the other hand, it has also stimulated research interest to develop new low-cost precursors and manufacturing process. Despite numerous studies, researchers have yet to develop a new polymeric precursor that is less expensive, melt-spinnable, and offers a high C-yield. In this three-year R&D project, we proposed three consecutive research phases, including (i) the development of new polyolefin-based precursors that can be thermally transformed to C material in one heating step under inert atmosphere and achieved a C-yield >80%, which is more than 60% higher than that of current PAN precursor, (ii) With the suitable new polyolefin-based precursor, we can focus on CF manufacturing process, including fiber spinning, stabilization, and carbonization, to form high quality carbon fibers with the desirable polymorphous morphology, and (iii) the investigation of thermal conversion process under tension for converting polyolefin-based precursor fibers to high tensile strength carbon fibers with specific graphite crystallite size, orientation of basal line, order-disorder ratio, fiber diameter, and reduced structural defect (voids). composites.

08 HYDROGEN↗
DOE OSTI2024

Intern Poster

Digital Image Correlation (DIC) is an optical technique that combines image registration and tracking methods for accurate two-dimensional and three-dimensional changes in images. DIC software can be used to track the contour, deformation, and strain of a sample. In the Advanced Test Reactor (ATR) at INL (Idaho National Laboratory) there exists a small working window of samples that can become irradiated. Hundreds of graphite disks called piggybacks have undergone this irradiation as part of the Advanced Reactor Technologies (ART) program. After irradiation, it is desirable to understand the change in tensile strength as a function of dose. Due to the limited space in the ATR, typical dog bone tensile tests reduce the number of graphite samples from hundreds to tens. However, there does exist an ASTM standard, D8289, which uses disc compression of graphite to estimate the tensile strength of the specimen with the Brazilian Disk test fixture. While only used as an estimate, which is typically off by a third, it is the purpose of this study to identify how to amend D8289 to remove the word "estimate" with the help of DIC.

36 MATERIALS SCIENCE↗
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 OSTI2023

Understanding the effect of specimen size on the properties of fine-grain isotropic nuclear graphite for irradiation studies: Mechanical properties

Programs at Oak Ridge National Laboratory are investigating the effects of neutron irradiation in the High Flux Isotope Reactor (HFIR) on properties of fine-grain isotropic nuclear graphites. Specimens were irradiated in the HFIR flux trap, which requires using specimens that are smaller than the sizes recommended by ASTM standards. Here, this work has investigated whether subsized specimens can be used to reliably predict strength behavior. The results show that (1) tensile strength was loosely coupled to specimen volume, (2) the compression specimens had a slightly higher strength for larger volumes, (3) the uniaxial flexural strength was nearly independent of specimen effective area for one orientation whereas the other orientation showed a slight strength–size scaling Weibull behavior, and (4) equibiaxial flexural strength followed traditional strength–size scaling Weibull behavior. These results suggest that fine-grain graphite is more similar to ceramics than to medium- and large-grained graphite with respect to both the strength-limiting flaws (Griffith theory) and the effect of machining on the strength behavior. This work shows that using specimens that are smaller than ASTM recommendations are suitable for investigating the strength changes caused by neutron irradiation because the primary interest is the relative change in the strength distribution parameters rather than their absolute values.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Mechanical behavior analysis using small-size graphite disc compression testing with digital image correlation methods

The ASTM International Standard Test Method for Tensile Stress-Strain of Carbon and Graphite (ASTM C749) requires dog-bone shaped specimens. Unfortunately, specimen sizes required in ASTM C749 are much too large for irradiation within a test reactor and are incompatible with irradiation capsule volumes, molten salt degradation facilities, or oxidation apparatus. The ASTM International Standard Test Method for Tensile Strength Estimate by Disc Compression of Manufactured Graphite (ASTM D8289) was developed as an alternative that provides a means for testing tensile properties using smaller specimens when geometric constraints are present. However, the important stress–strain relation throughout the test loading history that could be recorded using ASTM C749 is lost by using ASTM D8289 because of the difficulties involved with using traditional strain gauges or extensometers on such a small specimen. In this work, a digital image correlation (DIC) method, which uses a full-field noncontact surface displacement measurement technique, was applied along with the ASTM D8289 splitting tensile test on eight small samples of graphite grade Mersen 2114 to track their deformation during compressive loading. Initial results showed that the DIC technique can measure the surface displacement/strain on these small (Ø6 × 3 mm) graphite specimens with high accuracy and good repeatability, indicating that the DIC technique, combined with the small-disc splitting tensile test in ASTM D8289, can capture the strain evolution in the test. In conclusion, further comparison of measured strains with analytical and numerical simulation results provides additional understanding of the mechanical behaviour of this nuclear graphite material in complex stress states.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Heat-treated glassy carbon under pressure exhibiting superior hardness, strength and elasticity

Glassy carbon (GC) is a type of non-graphitizing disordered carbon material at ambient pressure and high temperatures, which has been widely used due to its excellent mechanical properties. Here we report the changes in the microstructure and mechanical properties of GC treated at high pressures (up to 5 GPa) and high temperatures. The formation of intermediate sp 2 –sp 3 phases is identified at moderate treatment temperatures before the complete graphitization of GC, by analyzing synchrotron X-ray diffraction, Raman spectra, and transmission electron microscopy images. The intermediate metastable carbon materials exhibit superior mechanical properties with hardness reaching up to 10 GPa and compressive strength reaching as high as 2.5 GPa, nearly doubling those of raw GC, and improving elasticity and thermal stability. The synthesis pressure used in this study can be achieved in the industry on a commercial scale, enabling the scalable synthesis of this type of strong, hard, and elastic carbon materials.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Mesophase pitch-based high performance carbon fiber production using coal extracts from mild direct coal liquefaction

Mild direct coal liquefaction (autogenous pressure, no catalyst, no H2 gas) of Springfield coal in fluid catalytic cracking decant oil is shown to effectively produce coal extract precursors to spinnable mesophase pitch. Here this work demonstrates that the coal extract can be thermally treated to obtain mesophase pitch in a facile one-step process, bypassing the production of an intermediate isotropic pitch. Furthermore, the presence of 25 wt.% coal in the initial slurry can increase the yield to mesophase pitch nearly twofold and yield to carbon fiber by approximately 70%. The coal extract-derived mesophase pitch was melt-spun and heat treated to produce carbon fiber with graphitic texture, high modulus (>400 GPa) and tensile strength up to 943 MPa. Overall, this work demonstrates that coal can be effectively utilized to markedly amplify the mesophase pitch and carbon fiber yield from fluid catalytic cracking decant oil by relatively simple processing, while conserving utility as a precursor to high performance carbon fiber and potentially other high value graphitic products.

36 MATERIALS SCIENCE↗
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 OSTI2024

Graphite Oxidation Activities

Graphite Oxidation Activities to include introduction, rate determination, strength after oxidation, penetration depth analysis, strategic partnership projects, and summary and continuing work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Technical Program Plan for INL Advanced Reactor Technologies Advanced Gas Reactor Fuel Development and Qualification Program

High-temperature gas cooled reactors (HTGRs) are graphite moderated nuclear reactors cooled with helium. Their high outlet temperatures and thermal energy conversion efficiency enable efficient and cost effective integration with non electricity generating applications. These applications include process heat and hydrogen production for petrochemical and other industrial processes that require operating temperatures between 300 and 900°C. HTGRs will supplement the use of premium fossil fuels such as oil and natural gas, improve overall energy security in the United States by reducing dependence on foreign fuels, and reduce carbon dioxide (CO2)/greenhouse gas emissions. The HTGR design uses helium as a coolant, graphite as a neutron moderator, and ceramic particle fuel. Helium is chemically inert and neutronically transparent. The graphite core slows down the neutrons, retains its strength at high temperatures, provides structural stability, and acts as a substantial heat sink during transient conditions. The ceramic particle fuel is extremely robust and retains the radioactive by products of the fission reaction within the coated particle under normal and off normal conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Technical Program Plan for INL Advanced Reactor Technologies Advanced Gas Reactor Fuel Development and Qualification Program

High-temperature gas cooled reactors (HTGRs) are graphite moderated nuclear reactors cooled with helium. Their high outlet temperatures and thermal energy conversion efficiency enable efficient and cost effective integration with non electricity generating applications. These applications include process heat and hydrogen production for petrochemical and other industrial processes that require operating temperatures between 300 and 900°C. HTGRs will supplement the use of premium fossil fuels such as oil and natural gas, improve overall energy security in the United States by reducing dependence on foreign fuels, and reduce carbon dioxide (CO2)/greenhouse gas emissions. The HTGR design uses helium as a coolant, graphite as a neutron moderator, and ceramic particle fuel. Helium is chemically inert and neutronically transparent. The graphite core slows down the neutrons, retains its strength at high temperatures, provides structural stability, and acts as a substantial heat sink during transient conditions. The ceramic particle fuel is extremely robust and retains the radioactive by products of the fission reaction within the coated particle under normal and off normal conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Technical Program Plan for INL Advanced Reactor Technologies Advanced Gas Reactor Fuel Development and Qualification Program

High-temperature gas cooled reactors (HTGRs) are graphite moderated nuclear reactors cooled with helium. Their high outlet temperatures and thermal energy conversion efficiency enable efficient and cost effective integration with non electricity generating applications. These applications include process heat and hydrogen production for petrochemical and other industrial processes that require operating temperatures between 300 and 900°C. HTGRs will supplement the use of premium fossil fuels such as oil and natural gas, improve overall energy security in the United States by reducing dependence on foreign fuels, and reduce carbon dioxide (CO2)/greenhouse gas emissions. The HTGR design uses helium as a coolant, graphite as a neutron moderator, and ceramic particle fuel. Helium is chemically inert and neutronically transparent. The graphite core slows down the neutrons, retains its strength at high temperatures, provides structural stability, and acts as a substantial heat sink during transient conditions. The ceramic particle fuel is extremely robust and retains the radioactive by products of the fission reaction within the coated particle under normal and off normal conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗