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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 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
NASA NTRS1991

Development of graphite/copper composites utilizing engineered interfaces

In situ measurements of graphite/copper alloy contact angles were made using the sessile drop method. The interfacial energy values obtained from these measurements were then applied to a model for the fiber matrix interfacial debonding phenomenon found in graphite/copper composites. The formation obtained from the sessile drop tests led to the development of a copper alloy that suitably wets graphite. Characterization of graphite/copper alloy interfaces subjected to elevated temperatures was conducted using Scanning Electron Microscopy, Energy Dispersive Spectroscopy, Auger Electron Spectroscopy, and X Ray Diffraction analyses. These analyses indicated that during sessile drop tests conducted at 1130 C for 1 hour, copper alloys containing greater than 0.98 at pct chromium form continuous reaction layers of approx. 10 microns in thickness. The reaction layers are adherent to the graphite surface. The copper wets the reaction layer to form a contact angle of 60 deg or less. X ray diffraction results indicate that the reaction layer is Cr3C2.

Devincent, Sandra M.↗
DOE OSTI2025

NDMAS

Overview of Current ART-GCR Data: Fuel Fabrication, Irradiation Monitoring (Fuel & Graphite – near real-time for HDG-1), Post-Irradiation Examination (Fuel & Graphite), Graphite Characterization (Baseline and Irradiated), High Temperature Metals Mechanical Tests, Design, Methods, and Validation Data, Japan Atomic Energy Agency’s High Temperature Test Reactor (HTTR), Argonne National Laboratory’s Natural convection Shutdown heat removal Test Facility (NSTF), Oregon State University’s High Temperature Test Facility (HTTF), Generation IV International VHTR Materials Handbook, Additional related data, and Advanced Test Reactor operations (near real-time).

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

AREVA Steam Cycle High Temperature Gas-Cooled Reactor Development and Deployment Plan - 18684

The development and deployment timeline for the AREVA SC-HTGR is among the shortest of all Generation IV reactor technologies. Once demonstrated, the SC-HTGR enters an expanding marketplace beyond electricity generation. This enables industrial growth that is today solely captive to fossil fuels, primarily natural gas supply. The SC-HTGR plant will be a hedge that insulates a wide range of process industries from fossil fuel price and supply volatility in coming decades when the SC-HTGR plant design is ready for deployment. Similar to existing light water reactors, the SC-HTGR is largely immune to fuel price swings for its expected 60 year operating life cycle. Approximately 70% of the SCHTGR plant energy cost is driven by the initial capital investment with fuel being less than 20%. This is entirely the opposite of natural gas fueled power plants where about 70% of the energy cost is directly tied to the cost and availability of fuel and the potential volatility this brings with it over a plant life cycle that could span several decades. This paper details the development and deployment strategy envisioned by AREVA for the SC-HTGR plant. The development phase includes completion of the full scale reactor module design and licensing, build and demonstrate the first commercial scale module, get design certification, and launch the commercial fleet deployment. The development plan takes full advantage of a) a large body of successful pre-application licensing interactions with the NRC under the DOE NGNP project, b) ongoing experiments and test results accumulated through the TRISO particle fuel qualification and nuclear grade graphite characterization R and D activities conducted in the U.S. and internationally, and c) operation of several small scale technology demonstration facilities and large scale commercial design projects. The abundance of the historic HTGR reactor technology demonstration projects provides a high degree of confidence that sufficient technical information is available to begin FOAK plant engineering and licensing followed by commercial demonstration of the first reactor module within a next decade. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Resolving Detonation NanoDiamond Size Evolution and Morphology at Sub-Microsecond Time-Scales During High-Explosive Detonations

Characterization of the initial morphology of detonation nanodiamond (DND) has been the focus of many research studies that aim to develop fundamental understanding of carbon condensation under extreme conditions. Identifying the pathways of DND formation has the potential for significant impact in many of the controlled synthesis of nanoscale carbon with tailored functionality; currently, a wide range of possible (and conflicting) mechanisms of nucleation and growth have been proposed and further research is essential. Building a comprehensive understanding of DND formation is challenging because it requires in situ characterization on the sub-µs time-scale during a high explosive detonation. In this study, time resolved small angle X-ray scattering (TR-SAXS) is used to reveal the early-stage DND morphology from < 0.1 μs to 6 μs after the detonation front passes through the X-ray beam path. We address the ambiguity of models previously reported for analysis of small angle scattering from DND by comparing (i) in situ, TR-SAXS recorded during early-stage particulate formation and (ii) ex situ SAXS and transmission electron microscopy measurements (TEM) of products recovered from detonation of the same high-explosive within a carefully designed ice chamber. The SAXS from both late-time ( > 1 μs) in situ and recovered DND exhibit consistent features in the I(q) curve. Such close similarity allows a high fidelity SAXS model derived from the ex situ SAXS and TEM measurements to be applied to the in situ data, which yields new insight into the early stage (< 1µs) morphology of DND. Our analysis indicates a size-distribution of DND particles is observed within 0.1 μs post-detonation and have a mean core diameter slightly below 4 nm and surface texture that consists of hemispherical protrusions that extend ~ 1 nm from the surface. Beyond ~ 0.3 μs, the size distribution increases slightly, while the surface texture persists for several microseconds after the detonation. Combined with thermochemical simulations, these results indicate that during detonation of Composition B, carbon is condensed into nanoscale diamond much faster than previously reported in other studies. Furthermore, the surface texture of the DND is shown to arise during condensation rather than via subsequent graphitization.

carbon condensation↗
DOE OSTI2024

NDMAS Portal Updates

Overview of current ART-GCR Data, including fuel fabrication, irradiation monitoring, fuel and graphite, post-irradiation examination, graphite characterization, high temperature metals mechanical tests, methods validation data (including: HTTR, NSTF, and HTTF), addition of AGR fuel data, other GCR and related data, updated high temperature metals, ongoing and upcoming work, users guide, and the link to those interested in access to NDMAS.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2010

Mode II Interlaminar Fracture Toughness and Fatigue Characterization of a Graphite Epoxy Composite Material

Mode II interlaminar fracture toughness and delamination onset and growth characterization data were generated for IM7/8552 graphite epoxy composite materials from two suppliers for use in fracture mechanics analyses. Both the fracture toughness testing and the fatigue testing were conducted using the End-notched Flexure (ENF) test. The ENF test for mode II fracture toughness is currently under review by ASTM as a potential standard test method. This current draft ASTM protocol was used as a guide to conduct the tests on the IM7/8552 material. This report summarizes the test approach, methods, procedures and results of this characterization effort.

O'Brien, T. Kevin↗
DOE OSTI2023

Microstructural characterization of nuclear graphite: from the microscale to the nanoscale

Multiple reactor designs use graphite as a moderator of the nuclear reactions and as structural support. During the lifetime of the reactor, multiple aging factors such as neutron irradiation, oxidation, and temperature along with others induce changes in the microstructure and crystal lattice of graphite components. The pore morphology and crystal structure of some phases in graphite can be used to trace the evolution of irradiation defects and mechanical properties of graphite. We present a combination of results from several microscopy techniques to investigate the differences between nuclear graphite grades and the effects of neutron irradiation and oxidation at multiple length scales. This multiscale approach is needed to understand the microstructural variations caused by the raw materials and manufacturing processes as well as how the different phases of graphite are affected by the reactor environment. The results provide insight into the oxidation- and radiation-induced changes of graphite and create a robust baseline of microstructure information that can be used for the selection of materials for the next generation of nuclear power stations. Moreover, the experiments conducted in this work provide an overview of the advantages and limitations of the most common techniques used to characterize nuclear graphite and how these techniques might be applied to study other carbon-based materials used in the nuclear industry.

Arregui Mena, Jose'↗
DOE OSTI2015

Characterization of nuclear graphite elastic properties using laser ultrasonic methods

Laser ultrasonic methods have been used to characterize the elastic behaviors of commercially-available and legacy nuclear graphites. Since ultrasonic techniques are sensitive to various aspects of graphite microstructure including preferred grain orientation, microcrack orientation and porosity, laser ultrasonics is a candidate technique for monitoring graphite degradation and structural integrity in environments expected in high-temperature, gas-cooled nuclear reactors. Aspects of materials texture can be assessed by studying ultrasonic wavespeeds as a function of propagation direction and polarization. Shear wave birefringence measurements, in particular, can be used to evaluate elastic anisotropy. In this work, laser ultrasonic measurements of graphite moduli have been made to provide insight into the relationship between the microstructures and the macroscopic stiffnesses of these materials. In particular, laser ultrasonic measurements have been made using laser line sources to produce shear waves with specific polarizations. By varying the line orientation relative to the sample, shear wave birefringence measurements have been recorded. Results from shear wave birefringence measurements show that an isostatically molded graphite, such as PCIB, behaves isotropically, while an extruded graphite, such as H-451, displays significant ultrasonic texture. Graphites have complicated microstructures that depend on the manufacturing processes used, and ultrasonic texture in these materials could originate from grain orientation and preferred microcrack alignment. Effects on material isotropy due to service related microstructural changes are possible and the ultimate aim of this work is to determine the degree to which these changes can be assessed nondestructively using laser ultrasonics measurements

anisotropy↗
DOE OSTI2014

Generation of graphite particles by sliding abrasion and their characterization

Characterization of graphite particles (dust) produced by abrasion that would occur in a pebble bed reactor is of interest for reasons of safety, operation, and maintenance. To better understand this abrasion and particle generation, we have built a test apparatus to produce particles by sliding abrasion in a 1% to 5% relative humidity air environment. We have used a commercial-grade graphite in our experiments and have generated size distributions for the abraded particles. We have also fit lognormal functions to those size distributions (for use in computer codes); determined particle shapes; measured temperature and humidity during the tests; measured and calculated wear rates; and measured the surface roughness of both pretest and posttest samples, particle surface areas, pore volumes, and pore volume distributions of particles produced during abrasion of graphite surfaces under different loadings and sliding speeds. The experiments showed that as loading (analogous to pebble depth in the reactor) and sliding speed increase, so do the wear rates and numbers of particles produced, while surface roughness decreases, increases, and then decreases. Brunauer-Emmett-Teller measurements show that abrasion increases surface area from 0.583 m 2 /g in the bulk material to 555 m 2 /g in material abraded at high loading and high sliding speed. Wear rates range from 0.005 to 0.991 g/m per contact site. As a result, the size of the particles observed was <4000 nm. Finally, in all, our research shows that pebble abrasion is a complex process that is not constant during operation and thus should be considered for future work.

36 MATERIALS SCIENCE↗
NASA NTRS2014

Computational Reduction of Specimen Noise to Enable Improved Thermography Characterization of Flaws in Graphite Polymer Composites

Flaw detection and characterization with thermographic techniques in graphite polymer composites are often limited by localized variations in the thermographic response. Variations in properties such as acceptable porosity, fiber volume content and surface polymer thickness result in variations in the thermal response that in general cause significant variations in the initial thermal response. These result in a "noise" floor that increases the difficulty of detecting and characterizing deeper flaws. A method is presented for computationally removing a significant amount of the "noise" from near surface porosity by diffusing the early time response, then subtracting it from subsequent responses. Simulations of the thermal response of a composite are utilized in defining the limitations of the technique. This method for reducing the data is shown to give considerable improvement characterizing both the size and depth of damage. Examples are shown for data acquired on specimens with fabricated delaminations and impact damage.

Winfree, William P.↗
DOE OSTI2015

Kelvin probe characterization of buried graphitic microchannels in single-crystal diamond

In this work, we present an investigation by Kelvin Probe Microscopy (KPM) of buried graphitic microchannels fabricated in single-crystal diamond by direct MeV ion microbeam writing. Metal deposition of variable-thickness masks was adopted to implant channels with emerging endpoints and high temperature annealing was performed in order to induce the graphitization of the highly-damaged buried region. When an electrical current was flowing through the biased buried channel, the structure was clearly evidenced by KPM maps of the electrical potential of the surface region overlying the channel at increasing distances from the grounded electrode. The KPM profiling shows regions of opposite contrast located at different distances from the endpoints of the channel. This effect is attributed to the different electrical conduction properties of the surface and of the buried graphitic layer. The model adopted to interpret these KPM maps and profiles proved to be suitable for the electronic characterization of buried conductive channels, providing a non-invasive method to measure the local resistivity with a micrometer resolution. The results demonstrate the potential of the technique as a powerful diagnostic tool to monitor the functionality of all-carbon graphite/diamond devices to be fabricated by MeV ion beam lithography.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2023

Report on initial development of a database of nuclear graphite characteristics based on microstructural characterization

This report outlines the current and future efforts to produce a comprehensive library of microstructures of nuclear graphite and carbon-based materials that are candidate materials for nuclear applications in the United States. This library must contain extensive characterizations of unirradiated graphite materials, a guide to some of the techniques used to characterize graphite, a compendium of characterization data of neutron-irradiated or oxidized material, and a compendium of microstructural information of carbon-based materials. These characterization efforts are being conducted at various length scales to understand these complex materials’ local structure and property relationships. Achieving this goal requires further developing or adapting advanced characterization techniques that capture graphite’s most relevant characteristics. Some of the general objectives of this project are to aid with the material selection, licensing, management, and core assessments of a graphite core by documenting the unirradiated microstructure of relevant grades or by characterizing the evolution of the microstructure under the reactor environment. Moreover, this project aims to provide additional information, guidelines for characterizing graphite, and a protocol to assess a nuclear graphite grade. This report also summarizes some of the initial results and some of the techniques commonly used to characterize nuclear graphite.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Machining Graphite Specimens from Sub- Blocks for Baseline Characterization

The Advanced Reactor Technologies (ART) Baseline Graphite Characterization program at the Idaho National Laboratory (INL) is performing a comprehensive evaluation of inter-billet variations in as-produced nuclear-grade graphite. To accomplish this, over 500 mechanical test specimens are machined from a single billet. A typical graphite billet is approximately 6 feet long by 2 feet in diameter or 2 ft square and weighs approximately 1500 pounds. Initially the full-size billets are pre-sectioned into sub-blocks that can be easily handled and shipped. In this SOW, 44 pre-labeled sub-blocks will be machined into mechanical test specimens by the supplier. These sub-blocks are supplied and shipped by the purchaser to the supplier.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Advanced Surface and Microstructural Characterization of Natural Graphite Anodes for Lithium Ion Batteries

Natural graphite powders were subjected to a series of thermal treatments in order to improve the anode irreversible capacity loss (ICL) and capacity retention during long-term cycling of lithium ion batteries. A baseline thermal treatment in inert Ar or N2 atmosphere was compared to cases with a proprietary additive to the furnace gas environment. This additive substantially altered the surface chemistry of the natural graphite powders and resulted in significantly improved long-term cycling performance of the lithium ion batteries over the commercial natural graphite baseline. Different heat-treatment temperatures were investigated ranging from 950-2900 C with the intent of achieving the desired long-term cycling performance with as low of a maximum temperature and thermal budget as possible. A detailed summary of the characterization data is also presented, which includes X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), Raman spectroscopy, and temperature-programed desorption mass spectroscopy (TPD-MS). This characterization data was correlated to the observed capacity fade improvements over the course of long-term cycling at high charge-discharge rates in full lithium-ion coin cells. It is believed that the long-term performance improvements are a result of forming a more stable solid electrolyte interface (SEI) layer on the anode graphite surfaces, which is directly related to the surface chemistry modifications imparted by the proprietary gas environment during thermal treatment.

Gallego, Nidia C↗
DOE OSTI2011

Carbon Characterization Laboratory Readiness to Receive Irradiated Graphite Samples

The Carbon Characterization Laboratory (CCL) is located in Labs C19 and C20 of the Idaho National Laboratory Research Center. The CCL was established under the Next Generation Nuclear Plant Project to support graphite and ceramic composite research and development activities. The research conducted in this laboratory will support the Advanced Graphite Creep experiments—a major series of material irradiation experiments within the Next Generation Nuclear Plant Graphite program. The CCL is designed to characterize and test low activated irradiated materials such as high purity graphite, carbon-carbon composites, silicon-carbide composite, and ceramic materials. The laboratory is fully capable of characterizing material properties for both irradiated and nonirradiated materials. Major infrastructural modifications were undertaken to support this new radiological facility at Idaho National Laboratory. Facility modifications are complete, equipment has been installed, radiological controls and operating procedures have been established and work management documents have been created to place the CCL in readiness to receive irradiated graphite samples.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

SIMS and HR-XPS characterization of lithiated graphite from the magnetic fusion device RFX-mod

Lithium wall conditioning has improved the performance of many magnetic fusion devices. Li conditioning in the RFX-mod device was performed by: (1) a single Li pellet injector and (2) a multi-Li pellet injector during He plasma discharges; (3) a Li evaporator after He glow discharge cleaning and He plasma discharge exposure. This report compares the spatial and depth distributions of Li deposited on polycrystalline graphite witness samples at different locations in RFX-mod and the elemental and chemical compositions of the resulting surfaces. The sample surfaces were analyzed ex situ using secondary ion mass spectrometry (SIMS) and high-resolution X-ray photoelectron spectroscopy (HR-XPS). The results showed that Li pellet injection provided a relatively uniform toroidal coverage while Li evaporation produced highly localized Li deposition. A Li 1s HR-XPS peak at 56.5 eV binding energy (BE) characteristic of lithium-intercalated graphite was only observed with the sample exposed to Li evaporation. All of the samples exhibited a HR-XPS C 1s peak at 285.1–285.2 eV BE that is largely attributed to hydrogenated graphite. This finding suggests that hydrogenation of fresh graphite occurs during He plasma discharge exposures. Finally, our results have implications for density control and the selection of Li conditioning techniques in magnetic fusion devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗