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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
NASA NTRS1980

Improved fiber retention by the use of fillers in graphite fiber/resin matrix composites

A variety of matrix fillers were tested for their ability to prevent loss of fiber from graphite fiber/PMR polyimide and graphite fiber/epoxy composites in a fire. The fillers tested included powders of boron, boron carbide lime glass, lead glass, and aluminum. Boron was the most effective and prevented any loss of graphite fiber during burning. Mechanical properties of composites containing boron filler were measured and compared to those of composites containing no filler.

Gluyas, R. E.↗
DOE OSTI2023

Nuclear fuel rods and heat pipes in a graphite moderator matrix for a micro-reactor, with the fuel rods having fuel pellets in a BeO sleeve

A reactor unit cell is disclosed including a graphite moderator structure, a heat pipe positioned in the graphite moderator structure, and a fuel assembly positioned in the graphite moderator structure. The fuel assembly comprises at least one fuel rod. Each fuel rod comprises a beryllium-oxide sleeve and nuclear fuel positioned in the beryllium-oxide sleeve.

Levinsky, Alex↗
DOE OSTI2023

Investigation of Coincidence Counting for Improving Minimal Detectable Activity of 110m Ag in Single Particle Gamma Analysis

Post-irradiation examination (PIE) of fuel particles from the fourth Advanced Gas Reactor Fuel Development and Qualification (AGR) Program irradiation (AGR-5/6/7) is being performed at Oak Ridge National Laboratory (ORNL). Tristructural isotropic (TRISO)-coated particles and associated compacts for the AGR-5/6/7 experiment fabricated by BWX Technologies Nuclear Operations Group were formed into a graphite matrix compact and irradiated at the Advanced Test Reactor at Idaho National Laboratory. At ORNL, particles are deconsolidated from the graphite matrix compact and individually scanned for emitted gamma rays with the Irradiated Microsphere Gamma Analyzer (IMGA). The IMGA system comprises a single high purity germanium (HPGe) detector, an automated particle handling vacuum system, and an ORTEC DSPEC-50 digital spectrometer for gamma ray analysis. IMGA quantifies gamma ray-emitting fission product inventories of individual TRISO particles, and these inventories can be compared with the measured average inventories per particle and radionuclide inventories predicted by AGR-5/6/7 physics calculations to determine if a particle experienced radionuclide release. Details on IMGA data collection methods can be found in the literature. The TRISO particle’s SiC layer provides structural support, as well as a barrier for fission product release during irradiation or subsequent safety testing. A weakened or compromised SiC layer can be identified by the release of radionuclides, such as 137 Cs, which is detected by IMGA. However, select radionuclides, such as 90 Sr, 110 mAg, and 154 Eu have been shown to migrate through an intact SiC layer. Measurement of the radionuclide 110 mAg is significant as its release has been shown to be particularly sensitive to in-reactor conditions (e.g., temperature) with broad variable particle to-particle release behaviors observed within a single compact. As such, 110 mAg activity is often used for particle selection for comprehensive PIE as bounding 110 mAg retention particles are hypothesized to represent limits in particle behaviors within a compact. As TRISO particle fuel PIE activities continue over time, IMGA measurements of the 110 mAg inventory are eventually hindered because of its relatively short half-life (~250 days). As the fuel ages from its end of irradiation (EOI) date, the measurement uncertainty and minimum detectable activity (MDA) of 110m Ag increase because the detector background continuum begins to dominate. For particles from the second AGR irradiation experiment (AGR-2), the 110m Ag MDA was above 20% of the calculated average particle inventory after approximately five half-lives, and 110m Ag activity was no longer measurable with IMGA after approximately seven half-lives. Therefore, coincidence counting approaches have been explored to determine feasibility of leveraging new approaches to overcome limitations associated with increasing MDA over time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Salt hydrate-based phase change thermal energy storage and encapsulation thereof

Among other things, the present disclosure relates to phase change material (PCM) composites composed of an PCM mixed with a nucleating agent contained within the pores of a graphite matrix and/or a hydrogel. The process to create these PCM composites includes coating the surface of graphite with a surfactant, compressing the graphite to form a matrix, then filling the graphite matrix with the PCM.

Odukomaiya, Adewale↗
DOE OSTI2023

Microstructural analysis of tristructural isotropic particles in high-temperature steam mixed gas atmospheres

High-temperature gas-cooled reactors (HTGRs) use tristructural isotropic (TRISO) particles embedded in a graphitic matrix material to form the integral fuel element. Potential off-normal reactor conditions for HTGRs include steam ingress with temperatures above 1,000 °C. Fuel element exposure to steam can cause the graphitic matrix material to evolve, forming an atmosphere composed of oxidants and oxidation products and potentially exposing the TRISO particles to these conditions. Investigating the oxidation response of TRISO particles exposed to a mixed gas atmosphere will provide insight into the stability under off-normal conditions. In this study, surrogate TRISO particles were exposed to high temperatures (T = 1,200 °C) in flowing steam (3% < pH 2 O < 21%) and CO (pCO < 1%) to determine the oxidation behavior of the SiC layer when exposed to various mixed gas atmospheres. Scanning electron microscopy, x-ray diffraction, and focused ion beam milling was used to determine the impact of CO and steam on the oxidation behavior of the SiC layer. Therefore, the data presented demonstrates how the SiC layer showed strong oxidation resistance due to limited SiO 2 growth and maintained its structural integrity under these off-normal conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Quantitative measurement of pebble fuel anisotropy using generalized ellipsometry

The renewed development of high-temperature gas-cooled reactors has created a need to modernize pebble fuel quality control methods. One important specification in these methods is the coefficient of thermal expansion (CTE) anisotropy, which reflects graphite grain alignment and affects fuel performance. This study investigates the feasibility of using two-modulator generalized ellipsometry microscopy (2-MGEM) as a potential surrogate metric for CTE anisotropy. Graphite matrix compacts with controlled levels of anisotropy were fabricated by varying pre-pressures during forming and were subsequently characterized using both optical dilatometry and 2-MGEM. Measured responses were compared to an isotropic reference pebble produced using an industrial manufacturing process to provide a practical baseline. The resulting measurements revealed a strong quantitative correlation between the CTE anisotropy ratio and the corresponding diattenuation ratio. However, propagated measurement uncertainties in the optical method were found to limit definitive pass/fail discrimination, highlighting the need for improved precision before qualification as a quality control tool. Still, the results establish a quantitative relationship between optical and CTE anisotropy and support further development of 2-MGEM as a potential surrogate method for evaluating graphite matrix anisotropy in pebble fuel manufacturing.

Raftery, Alicia [ORNL] (ORCID:0000000259076427)↗
NASA NTRS1985

Polymer matrix and graphite fiber interface study

Hercules AS4 graphite fiber, unsized, or with EPON 828, PVA, or polysulfone sizing, was combined with three different polymer matrices. These included Hercules 3501-6 epoxy, Hercules 4001 bismaleimide, and Hexcel F155 rubber toughened epoxy. Unidirectional composites in all twelve combinations were fabricated and tested in transverse tension and axial compression. Quasi-isotropic laminates were tested in axial tension and compression, flexure, interlaminar shear, and tensile impact. All tests were conducted at both room temperature, dry and elevated temperature, and wet conditions. Single fiber pullout testing was also performed. Extensive scanning electron microphotographs of fracture surfaces are included, along with photographs of single fiber pullout failures. Analytical/experimental correlations are presented, based on the results of a finite element micromechanics analysis. Correlations between matrix type, fiber sizing, hygrothermal environment, and loading mode are presented. Results indicate that the various composite properties were only moderately influenced by the fiber sizings utilized.

Adams, D. F.↗
DOE OSTI2021

Gamma Spectrometry Examination of the AGR-3/4 Irradiation

The results from gamma spectrometry examination of the different components from the combined third and fourth U.S. Advanced Gas Reactor (AGR) TRISO-coated particle fuel irradiation tests (AGR-3/4) have been analyzed. This experiment was designed to provide information about in-pile fission product migration. In each of the 12 capsules, a single stack of four compacts with designed-to-fail (DTF) particles surrounded by inner and outer graphite and/or graphitic matrix rings and a graphite sink ring were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). Gamma spectrometry has been used to evaluate the gamma-emitting fission product inventory of compacts from the irradiation and evaluate the burnup of these compacts based on the activity of the radioactive cesium isotopes (i.e., Cs-134 and Cs-137) in the compacts. Burnup from gamma spectrometry compares well with predicted burnup from simulations. The inner and outer rings were also examined by gamma spectrometry to evaluate the total fission product inventory of the rings, and gamma emission computed tomography (GECT) was used to investigate the spatial distribution of gamma-emitting fission products within the rings. This report focuses on results obtained via the Precision Gamma Scanner (PGS) for the compacts and inner and outer rings. These non-destructive gamma measurements are currently being compared to recent destructive measurements of spatial fission product distributions in the rings and to predictions made via mathematical modeling. A separate report discusses the entire AGR-3/4 fission product mass balance for all irradiation capsule components including the inner and outer rings, graphite sink rings, capsule spacers, capsule foils, through tubes, and felts. This report is similar to the conference proceedings: Harp, J.M., Demkowicz, P.A., and Stempien, J.D., “Initial gamma spectrometry examination of the AGR-3/4 irradiation,” International Topical Meeting High Temperature Reactor Technology (HTR 2016), Las Vegas, NV, USA, November 2016, Paper HTR2016-18593, but it has been updated to reflect all the data that has been collected on AGR-3/4 using PGS. The Appendices contain extensive data from the collected gamma spectra.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1981

Sub-T/g/ annealing studies of advanced epoxy-matrix graphite-fiber-reinforced composites

Annealing at temperatures below the glass transition temperature (sub-T/g/ annealing) in structural epoxy network glasses is found to affect long-term mechanical properties of symmetrically (+ or - 45 deg) reinforced 16-lamina graphite/epoxy composites. Post-cured specimens of Thornel 300 graphite/Narmco 5208 epoxy are sub-T(g) annealed at 140 C for about 10, 100, 1000, 10,000, and 100,000 min with a prior quenching above T(g). The ultimate tensile strength, strain-to-break, and static toughness of the composite materials are found to decrease as a function of sub-T(g) annealing time, while no weight loss is observed. Time-dependent changes in properties are explained by a physical aging concept based on free-volume decreases in the nonequilibrium glassy state of network epoxies, and results imply that changes in composite properties may occur with the service time.

Kong, E. S. W.↗
NASA NTRS1984

Standard test evaluation of graphite fiber/resin matrix composite materials for improved toughness

Programs sponsored by NASA with the commercial transport manufacturers to develop a technology data base are required to design and build composite wing and fuselage structures. To realize the full potential of composite structures in these strength critical designs, material systems having improved ductility and interlaminar toughness are being sought. To promote systematic evaluation of new materials, NASA and the commercial transport manufacturers have selected and standardized a set of five common tests. These tests evaluate open hole tension and compression performance, compression performance after impact at an energy level of 20 ft-lb, and resistance to delamination. Ten toughened resin matrix/graphite fiber composites were evaluated using this series of tests, and their performance is compared with a widely used composite system.

Chapman, Andrew J.↗
DOE OSTI2023

Structure–property relations in graphitic pebbles for nuclear applications

This work presents an analytical approach for holistically characterizing graphitic matrix pebbles for nuclear applications whereby the macrostructure, microstructure, and thermophysical properties of pebbles are determined. A systematic sectioning method was applied to several pebbles to describe the regional properties of the samples. Intact matrix-only spheres and sections of spheres fabricated by Kairos Power were characterized via optical imaging, x-ray computed tomography, x-ray diffraction, and ellipsometry to determine 2D and 3D macrostructure and anisotropy. The thermophysical properties of these materials were determined via measurements of density, specific heat, thermal expansion, and thermal diffusivity. The results of this study indicate that the pebble fabrication methods and their resultant effect on microstructure have a nontrivial effect on thermophysical properties, confirming the importance of robust characterization of these components. A discussion of the characterization approach and its applicability to nuclear fuel development activities is also included.

42 ENGINEERING↗
NASA NTRS1972

Composites - Lighter and cheaper.

Composite engineering materials have two distinct parts, including the load-bearing filaments and the matrix. Fibers can range from glass to boron and graphite. Matrix constituents include epoxies for lower temperatures and metals for higher temperatures. Composites appear as tape or combined with metal in structural shapes. The increased use of composites will reduce over-all per pound manufacturing costs. Two major deterrents to further application of composites in advanced commercial transports include a shortage of cost-effective structural design concepts and a lack of confidence in the life characteristics. A program has been outlined by NASA that will, if implemented, provide considerable in-service life experience with structural components on commercial transports.

Goble, R. L.↗
DOE OSTI2021

AGR-5/6/7 Fuel Fabrication Report

The U.S. Department of Energy Office of Nuclear Energy (DOE NE) and the Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification program (referred to as AGR Fuel program hereafter) are pursuing qualification of tristructural isotropic (TRISO) coated particle fuel for use in high temperature gas cooled reactors (HTGRs). The AGR Fuel program was established to provide a fuel qualification data set in support of the licensing and operation of an HTGR. BWX Technologies Nuclear Operations Group (BWXT-NOG) was subcontracted to fabricate the fuel for the AGR program. Several investments and innovations were realized in preparation to fabricate fuel for the AGR-5/6/7 irradiation experiments that brought fuel fabrication fully out of the laboratory and into engineering-scale operations. These included: • Increased the kernel fabrication line capacity and uniformity • Upgraded ancillary support equipment and processes for the tristructural isotropic (TRISO) coating furnace • Demonstrated efficient production of the matrix precursor powder by dry jet milling of co mingled components • Demonstrated an engineering-scale method for quick and efficient overcoating TRISO particles with the matrix precursor • Demonstrated an automated, multi cavity compacting system with a volumetric feed system • Demonstrated a combined-cycle thermal treatment furnace These changes increased production rates of some of these processes by an order of magnitude or more while eliminating the use of flammable solvents, multiple grinding and sorting operations, and the weighing out of individual die charges. Three fuel kernel lots were fabricated for production of the fuel for AGR-5/6/7. The initial lot (J52R-16-39316) was certified to fuel specifications but was not used because the kernels had a high fraction of internal fissures that caused an unacceptable fraction of the kernels to fragment when charged to the coating furnace where the TRISO coating would be deposited. Fragmented kernels increased the dispersed uranium in the particles and produced a worrisome fraction of dimpled particles with an elevated probability of in-pile failure. After some efforts to identify the cause of the fissure formation, two additional lots were produced with much lower fissure fractions, J52R-16-69317 and 69318. The latter kernel lot was a backup to the first and was not needed. Multiple kernel batches were composited to form each of the lots so as to simulate a commercial-scale operation where kernel batches would also be composited. Multiple TRISO coating runs were performed and the product characterized so that several could be composited into a TRISO lot. TRISO lot J52R-16-98005 conformed to all fuel specifications except the mean outer pyrocarbon (OPyC) layer thickness was thinner than specified. Furthermore, it was determined that the TRISO lot had a dispersed uranium fraction (DUF) that might result in the compacts not meeting the DUF specification. A review of the role of the OPyC layer and consequences of the DUF by the Technical Coordination Team and INL concluded that the fuel was acceptable for use in the AGR-5/6/7 irradiation experiment. The TRISO particles were overcoated with the matrix precursor that had been produced in a jet-milling operation. The overcoating was performed in equipment originally designed to coat pharmaceuticals. The overcoating process performed well; producing highly spherical and uniform overcoats requiring little upgrading and no recycle or rework. TRISO particles were overcoated with the matrix precursor to achieve nominal volumetric packing fractions (PFs) of TRISO particles of 25% and 40% for the irradiation experiments. The 40% PF compacts occupy the first and fifth test capsule in the test train while the inner three capsules are loaded with 25% PF compacts. The resinated graphite matrix precursor powder was a derivative of the German A3-27 matrix formulation, which differs from previous AGR irradiation campaigns that used an A3-3 formulation. Jet milling of the matrix powder precursor produced a finer mean graphite particle size than the milling operations used for the A3-3 matrix powder precursor. Changes made in the matrix formula and equipment yielded compacts with significantly higher matrix density than was attained in previous AGR irradiation campaigns. The changes in the matrix formulation and the means of milling the powders also complicated resolution of the three fuel compact defect fractions, DUF, exposed kernel fraction (EKF), and the silicon carbide defect fraction (SDF). Characterization data from BWXT-NOG had some anomalous results, so samples of the fuel compacts and overcoated TRISO particles were also analyzed by Oak Ridge National Laboratory (ORNL) to ensure that the defect fractions were accurately characterized.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2004

Resistivity of Carbon-Carbon Composites Halved

Carbon-carbon composites have become the material of choice for applications requiring strength and stiffness at very high temperatures (above 2000 C). These composites comprise carbon or graphite fibers embedded in a carbonized or graphitized matrix. In some applications, such as shielding sensitive electronics in very high temperature environments, the performance of these materials would be improved by lowering their electrical resistivity. One method to lower the resistivity of the composites is to lower the resistivity of the graphite fibers, and a proven method to accomplish that is intercalation. Intercalation is the insertion of guest atoms or molecules into a host lattice. In this study the host fibers were highly graphitic pitch-based graphite fibers, or vapor-grown carbon fibers (VGCF), and the intercalate was bromine. Intercalation compounds of graphite are generally thought of as being only metastable, but it has been shown that the residual bromine graphite fiber intercalation compound is remarkably stable, resisting decomposition even at temperatures at least as high as 1000 C. The focus of this work was to fabricate composite preforms, determine whether the fibers they were made from were still intercalated with bromine after processing, and determine the effect on composite resistivity. It was not expected that the resistivity would be lowered as dramatically as with graphite polymer composites because the matrix itself would be much more conductive, but it was hoped that the gains would be substantial enough to warrant its use in high-performance applications. In a collaborative effort supporting a Space Act Agreement between the NASA Glenn Research Center and Applied Sciences, Inc. (Cedarville, OH), laminar preforms were fabricated with pristine and bromine-intercalated pitch-based fibers (P100 and P100-Br) and VGCF (Pyro I and Pyro I-Br). The green preforms were carbonized at 1000 C and then heat treated to 3000 C. To determine whether the fibers in the samples were still intercalated after composite fabrication, they were subjected to X-ray diffraction. The composites containing intercalated graphite fibers showed much higher background scatter than that of pristine fibers, indicating the presence of bromine in the samples. More importantly, faint features indicative of intercalation were visible in the diffraction pattern, showing that the fibers were still intercalated.

Gaier, James R.↗
DOE OSTI2023

A neutron tomography study to visualize fluoride salt (FLiNaK) intrusion in nuclear-grade graphite

Manufactured graphite is a preferred material for in-core components of molten salt reactors and fluoride salt-cooled high-temperature reactors, which are in permanent contact with liquid salts. However, owing to the porous nature of nuclear graphite, under certain conditions, molten salts may intrude graphite's pores and affect graphite's properties and functionality. Therefore, a better understanding of molten salt intrusion (distribution across sample cross section and penetration depth) is needed to assess its effects. Here, in this work, we have demonstrated the use of neutron imaging (computed tomography) in the evaluation of salt penetration and distribution of a wide range of graphite grades with diverse microstructures that have been subjected to FLiNaK (LiF–NaF–KF) intrusion at 750 °C and 5 bar pressure for 12 h. Because of the great neutron attenuation contrast from scattering and adsorption between Li (from FLiNaK) and the graphite matrix, we have obtained direct visualization of FLiNaK salt distribution in the salt-impregnated graphites for the first time. Three-dimensional reconstructed images and cross-sectional concentration profiles demonstrate that salt penetration and density distribution are greatly dependent on the microstructural properties of the graphite grade.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1977

Research on graphite reinforced glass matrix composites

The results of research for the origination of graphite-fiber reinforced glass matrix composites are presented. The method selected to form the composites consisted of pulling the graphite fiber through a slurry containing powdered glass, winding up the graphite fiber and the glass it picks up on a drum, drying, cutting into segments, loading the tape segment into a graphite die, and hot pressing. During the course of the work, composites were made with a variety of graphite fibers in a glass matrix.

Bacon, J. F.↗
NASA NTRS1991

Brief review of Rover fuel development at Los Alamos

A brief review of the graphite matrix uranium fuel development efforts at Los Alamos from 1955 through 1972 is presented. The uses of graphite flour carbon black, various binders, uranium dioxide, coated UC2 particles, and zirconium carbide in this development are described.

Davidson, Keith V.↗