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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 415 records · Page 23
All sourcesNASA NTRSDOE OSTI
NASA NTRS1995

Effects of stitching on fracture toughness of uniweave textile graphite/epoxy laminates

The effects of through-the-thickness stitching on impact damage resistance, impact damage tolerance, and Mode 1 and Mode 2 fracture toughness of textile graphite/epoxy laminates were studied experimentally. Graphite/epoxy laminates were fabricated from AS4 graphite uniweave textiles and 3501-6 epoxy using Resin Transfer Molding. The cloths were stitched with Kevlar(tm) and glass yarns before resin infusion. Delamination was implanted during processing to simulate impact damage. Sublaminate buckling tests were performed in a novel fixture to measure Compression After Impact (CAI) strength of stitched laminates. The results show that CAI strength can be improved up to 400% by through-the-thickness stitching. Double Cantilever Beam tests were performed to study the effect of stitching on Mode 1 fracture toughness G(sub 1c). It was found that G(sub 1c) increased 30 times for a low stitching density of 16 stitches/sq in. Mode 2 fracture toughness was measured by testing the stitched beams in End Notch Flexure tests. Unlike in the unstitiched beams, crack propagation in the stitched beams was steady. The current formulas for ENF tests were not found suitable for determining G(sub 2C) for stitched beams. Hence two new methods were developed - one based on crack area measured from ultrasonic C-scanning and the other based on equivalent crack area measured from the residual stiffness of the specimen. The G(sub 2c) was found to be at least 5-15 times higher for the stitched laminates. The mechanisms by which stitching increases the CAI strength and fracture toughness are discussed.

Sankar, Bhavani V.↗
DOE OSTI2014

Perfluoro Aryl Boronic Esters as Chemical Shuttle Additives in Lithium Ion Batteries

Redox shuttles could provide an added measure of control for lithium ion batteries beyond that which is provided by the battery management system. Redox shuttles, which are electrolyte additives, provide electrochemical overcharge protection. Since they possess a reduction-oxidation potential that is slightly above the maximum cathode potential during normal use, they should cause no degradation in cell performance during normal operation, but should shuttle excess electrons during overcharge conditions, effectively limiting the full cell voltage. The goals of the Chemical Shuttle Additives program were to discover and implement a redox shuttle that is compatible with large format lithium ion cells utilizing LiNi 1/3 Mn 1/3 Co 1/3 O 2 (NMC) cathode material and to understand the mechanism of redox shuttle action. Many redox shuttles, both commercially available and experimental, were tested and much fundamental information regarding the mechanism of redox shuttle action was discovered. In particular, studies surrounding the mechanism of the reduction of the oxidized redox shuttle at the carbon anode surface were particularly revealing. Shuttles that were evaluated in various lithium ion cell chemistries include: • 2-(pentafluorophenyl)-tetrafluoro-1,3,2-benzodioxaborole (BDB) supplied by Argonne National Laboratory (ANL), Lemont, Illinois; • 2,5-di-tert-butyl-1,4-dimethoxybenzene (DDB) supplied by 3M, St. Paul, Minnesota; • Li 2 B 12 F 12 supplied by Air Products, Allentown, Pennsylvania (pure compound) and Showa Denko, Japan (in electrolyte); • 1,4-bis(2-methoxyethoxy)-2,5-di-tert-butyl-benzene (ANL-RS2) supplied by Argonne National Laboratory (ANL), Lemont, Illinois; and • A 4.5V class redox shuttle supplied by Argonne National Laboratory (ANL), Lemont, Illinois Although redox shuttles with an appropriate redox potential and sufficient chemical and electrochemical stability for cells utilizing cathodes with potentials higher and lower than that of NMC, a redox shuttle that is viable for use in large format lithium ion cells with NMC cathodes was not found. The effect of the anode on the capacity loss of lithium ion cells with redox shuttles in cells utilizing lithium, lithium titanate, or graphite cells was studied. Molecular imprinting of the redox shuttle molecule during solid electrolyte interphase (SEI) layer formation likely contributes to the successful reduction of oxidized redox shuttle species at carbon anodes. This helps understand how a carbon anode covered with an SEI layer, that is supposed to be electrically insulating, can reduce the oxidized form of a redox shuttle.

25 ENERGY STORAGE↗
DOE OSTI2020

Investigating the Effects of Lithium Deposition on the Abuse Response of Lithium-Ion Batteries

Li deposition at the graphitic anode is widely reported in literature as one of the leading causes of capacity fade in lithium-ion batteries (LIBs). Previous literature has linked Li deposition resulting from low-temperature ageing to diminished safety characteristics, however no current research has probed the effects of Li deposition on the abuse response of well-characterized cells. Using overtemperature testing, a relationship between increased concentrations of Li deposition and exacerbated abuse response in 1 Ah pouch cells has been established. A novel Li deposition technique is also investigated, where cells with n:p < 1 (anode-limiting) have been cycled at a high rate to exploit Li + diffusion limitations at the anode. Scanning Electron Microscopy of harvested anodes indicates substantial Li deposition in low n:p cells after 20 cycles, with intricate networks of Li(s) deposits which hinder Li + intercalation/deintercalation. Peak broadening and decreased amplitude of differential capacity plots further validates a loss of lithium inventory to Li + dissolution, and Powder X-ray Diffraction indicates Li + intercalation with staging in anode interstitial sites as the extent of Li deposition increases. A cradle-to-grave approach is leveraged on cell fabrication and testing to eliminate uncertainty involving the effects of cell additives on Li deposition and other degradation mechanisms.

25 ENERGY STORAGE↗
DOE OSTI1991

Nuclear Safety [Vol. 32, No. 4, October-December 1991]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 477 Report on the International Symposium on the Use of Probabilistic Safety Assessment for Operational Safety—PSA '91, S. Chakraborty and M. Khatib-Rahbar; 488 Good Relationships Are Pivotal in Nuclear Data Bases, A. S. Heger and B. V. Koen; 494 Technical Note: The Interagency Nuclear Safety Review Panel's Evaluation of the Ulysses Space Mission, J. A. Sholtis, Jr., D. A. Huff, L. B. Gray, N. P. Klug, and R. O. Winchester; ACCIDENT ANALYSIS: 502 The Severe Accident Analysis Program for the Savannah River Nuclear Production Reactors, M. L. Hyder; CONTROL AND INSTRUMENTATION: 511 A Framework for Selecting Suitable Control Technologies for Nuclear Power Plant Systems, R. A. Kisner; DESIGN FEATURES: 521 Containments for Gas-Cooled Power Reactors History and Status, P.W. Williams; ENVIRONMENTAL EFFECTS: 537 Indoor Radon: A Natural Risk, N. H. Harley and J.H. Harley; On the Importance of the Atmospheric Parameters in the Fission Products Distribution of a Severe Reactor Accident, M. C. Barla and A. R. Bayulken; Book Review: Health Effects of Exposure to Low Levels of Ionizing Radiation BEIR V, C. R. Richmond; WASTE AND SPENT FUEL MANAGEMENT: 555 Activities Related to Waste and Spent Fuel Management, M. D. Muhlheim and E. G. Silver OPERATING EXPERIENCES: 567 Effects of Component Aging on the Westhinghouse Control Rod Drive System, K. Sullivan and W. Gunther; 577 Reactor Shutdown Experience, Compiled by J. W. Cletcher, 580 Selected Safety-Related Events, Compiled by G. A. Murphy; 582 Operating U 8 Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver, RECENT DEVELOPMENTS: 596 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 610 Reports, Standards, and Safety Guides, D. S. Queener; 614 Proposed Rule Changes as of June 30, 1991; ANNOUNCEMENTS: 520 Workshop on PC PRAISE: A Probabilistic Fracture Mechanics Personal Computer Code for Nuclear Power Plant Piping Reliability Assessment; 536 Fifth Workshop on Nuclear Power Plant Containment Integrity; 624 New OECD "International Information System on Occupational Exposure (ISOE)"; 625 Harvard School of Public Health Announces Short Courses; 625 Eighth Power Plant Dynamics, Control and Testing Symposium; 626 Second Training Course on Off-Site Emergency Planning and Response for Nuclear Accidents; 618 The Authors; 622 Reviewers of Nuclear Safety, Vol 32.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2004

Pulmonary toxicity of single-wall carbon nanotubes in mice 7 and 90 days after intratracheal instillation

Nanomaterials are part of an industrial revolution to develop lightweight but strong materials for a variety of purposes. Single-wall carbon nanotubes are an important member of this class of materials. They structurally resemble rolled-up graphite sheets, usually with one end capped; individually they are about 1 nm in diameter and several microns long, but they often pack tightly together to form rods or ropes of microscopic sizes. Carbon nanotubes possess unique electrical, mechanical, and thermal properties and have many potential applications in the electronics, computer, and aerospace industries. Unprocessed nanotubes are very light and could become airborne and potentially reach the lungs. Because the toxicity of nanotubes in the lung is not known, their pulmonary toxicity was investigated. The three products studied were made by different methods and contained different types and amounts of residual catalytic metals. Mice were intratracheally instilled with 0, 0.1, or 0.5 mg of carbon nanotubes, a carbon black negative control, or a quartz positive control and euthanized 7 d or 90 d after the single treatment for histopathological study of the lungs. All nanotube products induced dose-dependent epithelioid granulomas and, in some cases, interstitial inflammation in the animals of the 7-d groups. These lesions persisted and were more pronounced in the 90-d groups; the lungs of some animals also revealed peribronchial inflammation and necrosis that had extended into the alveolar septa. The lungs of mice treated with carbon black were normal, whereas those treated with high-dose quartz revealed mild to moderate inflammation. These results show that, for the test conditions described here and on an equal-weight basis, if carbon nanotubes reach the lungs, they are much more toxic than carbon black and can be more toxic than quartz, which is considered a serious occupational health hazard in chronic inhalation exposures.

Nanotubes, Carbon/adverse effects↗
DOE OSTI2016

Mixed effects associated with air ingress in graphite channels

The next generation of nuclear power plants, such as Modular High Temperature Gas-Cooled Reactors (MHTGR), have passive safety design features to overcome transients during off-normal operation but it is unproven whether their safety systems will perform as per design requirements. The design and construction of a facility, to investigate the passive safety response of the complete reactor system will require significant resources. Therefore, a practical approach is to improve the understanding of this multi-physics or mixed effect problem in laboratory scale set-ups and computational models. The HTGR related separate effects experiments at academic institutions and other laboratories have been developed to provide validation data and complement computational efforts to identify critical issues related to passive safety design. There are several separate effect experimental studies currently under investigation or recently conducted by several academic institutions in U.S and throughout the world. The High Temperature Test Facility (HTTF) has been constructed at Oregon State University to conduct integral thermal-hydraulics experiments for the development of HTGRs. For Depressurized Conduction Cooling (DCC) experiments in HTTF, a mixture of nitrogen and helium gas is used to study internal natural convection up to 650?C. Similarly many other experimental studies have emphasized on HTGR core and plenum-to-plenum thermal-hydraulic behavior but with inert gases such as nitrogen or non-oxidizing solid substitutes for graphite. There is limited experimental data of mixed oxidation effects upon air ingress in nuclear grade graphite leading changes to thermal properties, heat generation due to chemical reactions and natural convection of air-helium mixture inside core at high temperatures (above 900?C). This work describes a unique experimental set-up developed by authors to conduct these mixed effect experiments. The current work is equally relevant to the designs well known by other names or geometries such as Very High Temperature Reactors (VHTRs) and Pebble Bed Modular Reactors (PBMRs). A common feature of HTGRs with passive safety design is that under certain accident conditions the decay heat is dissipated passively from the core by radiation/conduction/convection heat transport to the surrounding environment. Passive decay heat removal is one of the prime passive safety features of HTGRs and the heat conduction through the internal core region, graphite matrix, to the periphery is one of the major mechanisms of passive safety in HTGRs. The conduction and radiation heat transfer provide long-term heat removal but, during a Depressurized Loss of Forced Coolant (DLOFC) accident, the anticipated break in the coolant system will allow air ingress into the hot core region and can lead to graphite oxidation, which will negatively impact the core thermal behavior. In addition, air will also impact natural circulation of Helium and thus diminish the heat removal rate, which can further increase the graphite temperature. Therefore, both air ingress can eventually give rise to thermal instabilities and then to possible situations for breaching the safety envelope of TRISO fuel. This research summary is organized to present the background on oxidation studies, description of experimental setup and discussion of results in the following sections. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

AGC-2 Irradiation Report

The Next Generation Nuclear Plant (NGNP) will be a helium-cooled, very high temperature reactor (VHTR) with a large graphite core. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs.[ , ] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphites have been developed and are considered suitable candidates for the new NGNP reactor design. To support the design and licensing of NGNP core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade with a specific emphasis on data related to the life limiting effects of irradiation creep on key physical properties of the NGNP candidate graphites. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change. In this case, the graphite physically changes i.e., first shrinking and then expanding with increasing neutron dose. This disparity in material volume change can induce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the components longer useful lifetimes within the core. Determining the irradiation creep rates of nuclear grade graphites is critical for determining the useful lifetime of graphite components and is a major component of the Advanced Graphite Creep (AGC) experiment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2001

Transient Reliability Analysis Capability Developed for CARES/Life

The CARES/Life software developed at the NASA Glenn Research Center provides a general-purpose design tool that predicts the probability of the failure of a ceramic component as a function of its time in service. This award-winning software has been widely used by U.S. industry to establish the reliability and life of a brittle material (e.g., ceramic, intermetallic, and graphite) structures in a wide variety of 21st century applications.Present capabilities of the NASA CARES/Life code include probabilistic life prediction of ceramic components subjected to fast fracture, slow crack growth (stress corrosion), and cyclic fatigue failure modes. Currently, this code can compute the time-dependent reliability of ceramic structures subjected to simple time-dependent loading. For example, in slow crack growth failure conditions CARES/Life can handle sustained and linearly increasing time-dependent loads, whereas in cyclic fatigue applications various types of repetitive constant-amplitude loads can be accounted for. However, in real applications applied loads are rarely that simple but vary with time in more complex ways such as engine startup, shutdown, and dynamic and vibrational loads. In addition, when a given component is subjected to transient environmental and or thermal conditions, the material properties also vary with time. A methodology has now been developed to allow the CARES/Life computer code to perform reliability analysis of ceramic components undergoing transient thermal and mechanical loading. This means that CARES/Life will be able to analyze finite element models of ceramic components that simulate dynamic engine operating conditions. The methodology developed is generalized to account for material property variation (on strength distribution and fatigue) as a function of temperature. This allows CARES/Life to analyze components undergoing rapid temperature change in other words, components undergoing thermal shock. In addition, the capability has been developed to perform reliability analysis for components that undergo proof testing involving transient loads. This methodology was developed for environmentally assisted crack growth (crack growth as a function of time and loading), but it will be extended to account for cyclic fatigue (crack growth as a function of load cycles) as well.

Nemeth, Noel N.↗
NASA NTRS2012

NASA Tech Briefs, May 2012

Topics covered include: An "Inefficient Fin" Non-Dimensional Parameter to Measure Gas Temperatures Efficiently; On-Wafer Measurement of a Multi-Stage MMIC Amplifier with 10 dB of Gain at 475 GHz; Software to Control and Monitor Gas Streams; Miniaturized Laser Heterodyne Radiometer (LHR) for Measurements of Greenhouse Gases in the Atmospheric Column; Anomaly Detection in Test Equipment via Sliding Mode Observers; Absolute Position of Targets Measured Through a Chamber Window Using Lidar Metrology Systems; Goldstone Solar System Radar Waveform Generator; Fast and Adaptive Lossless Onboard Hyperspectral Data Compression System; Iridium Interfacial Stack - IrIS; Downsampling Photodetector Array with Windowing; Optical Phase Recovery and Locking in a PPM Laser Communication Link; High-Speed Edge-Detecting Line Scan Smart Camera; Optical Communications Channel Combiner; Development of Thermal Infrared Sensor to Supplement Operational Land Imager; Amplitude-Stabilized Oscillator for a Capacitance-Probe Electrometer; Automated Performance Characterization of DSN System Frequency Stability Using Spacecraft Tracking Data; Histogrammatic Method for Determining Relative Abundance of Input Gas Pulse; Predictive Sea State Estimation for Automated Ride Control and Handling - PSSEARCH; LEGION: Lightweight Expandable Group of Independently Operating Nodes; Real-Time Projection to Verify Plan Success During Execution; Automated Performance Characterization of DSN System Frequency Stability Using Spacecraft Tracking Data; Web-Based Customizable Viewer for Mars Network Overflight Opportunities; Fabrication of a Cryogenic Terahertz Emitter for Bolometer Focal Plane Calibrations; Fabrication of an Absorber-Coupled MKID Detector; Graphene Transparent Conductive Electrodes for Next- Generation Microshutter Arrays; Method of Bonding Optical Elements with Near-Zero Displacement; Free-Mass and Interface Configurations of Hammering Mechanisms; Wavefront Compensation Segmented Mirror Sensing and Control; Long-Life, Lightweight, Multi-Roller Traction Drives for Planetary Vehicle Surface Exploration; Reliable Optical Pump Architecture for Highly Coherent Lasers Used in Space Metrology Applications; Electrochemical Ultracapacitors Using Graphitic Nanostacks; Improved Whole-Blood-Staining Device; Monitoring Location and Angular Orientation of a Pill; Molecular Technique to Reduce PCR Bias for Deeper Understanding of Microbial Diversity; Laser Ablation Electrodynamic Ion Funnel for In Situ Mass Spectrometry on Mars; High-Altitude MMIC Sounding Radiometer for the Global Hawk Unmanned Aerial Vehicle; PRTs and Their Bonding for Long-Duration, Extreme-Temperature Environments; Mid- and Long-IR Broadband Quantum Well Photodetector; 3D Display Using Conjugated Multiband Bandpass Filters; Real-Time, Non-Intrusive Detection of Liquid Nitrogen in Liquid Oxygen at High Pressure and High Flow; Method to Enhance the Operation of an Optical Inspection Instrument Using Spatial Light Modulators; Dual-Compartment Inflatable Suitlock; Large-Strain Transparent Magnetoactive Polymer Nanocomposites; Thermodynamic Vent System for an On-Orbit Cryogenic Reaction Control Engine; Time Distribution Using SpaceWire in the SCaN Testbed on ISS; and Techniques for Solution- Assisted Optical Contacting.

Source record↗
NASA NTRSDate not supplied

Production and Characterization of Additively Manufactured Radiator Panels with Integral Branching Heat Pipes for High-Temperature Heat Rejection

Emerging concepts for fission surface power and nuclear electric propulsion necessitate lightweight, mechanically robust, and thermally efficient heat rejection radiators. State-of-the-art intermediate-temperature (~400 K) composite radiator assemblies have been developed based on titanium-water heat pipes bonded to metal, graphite, and carbon-fiber-based panels. NASA has identified a need for new radiator concepts that can operate at even higher temperatures (500 – 600 K), minimize thermal resistances and thermal stress failures at bond interfaces, and approach areal densities of 2 – 3 kg m -2 . To meet these needs, our team is developing additively manufactured (AM) radiator panels with integral branching wicking heat pipe networks. Water is selected as the working fluid for this temperature range. Based on simulations and thermal vacuum experiments, these branching embedded heat pipe networks can efficiently distribute heat over panels for finned surface efficiencies of η f >70% at T H = 500 K input heat. This paper first presents laser powder-bed fusion AM strategies to produce embedded porous structures for wicking heat pipes in Inconel 718 and titanium alloys (commercially pure and Ti-6Al-4V alloys). Post-build chemical and thermal treatments are described that yield hydrophilic wicking surfaces for operation with water. Transient rate-of-rise experiments with water and acetone are reported that yield estimates for AM wick porosity (ϵ), permeability (K), and effective pore radius (r pore ). Based on the wick characterization results, small prototype radiator panels (75 × 125 mm) with integrated heat pipe networks were manufactured. Heat rejection performance data are presented from cold thermal vacuum testing, with heat input temperatures up to ~510 K. Future efforts will focus on improving heat pipe performance, optimizing radiator mass, and evaluating larger panels to assess scalability.

thermal management↗
NASA NTRS2024

Production and Characterization of Additively Manufactured Radiator Panels With Integral Branching Heat Pipes for High-Temperature Heat Rejection

Emerging concepts for fission surface power and nuclear electric propulsion necessitate lightweight, mechanically robust, and thermally efficient heat rejection radiators. State-of-the-art intermediate-temperature (~400 K) composite radiator assemblies have been developed based on titanium-water heat pipes bonded to metal, graphite, and carbon-fiber-based panels. NASA has identified a need for new radiator concepts that can operate at even higher temperatures (500 – 600 K), minimize thermal resistances and thermal stress failures at bond interfaces, and approach areal densities of 2 – 3 kg m -2 . To meet these needs, our team is developing additively manufactured (AM) radiator panels with integral branching wicking heat pipe networks. Water is selected as the working fluid for this temperature range. Based on simulations and thermal vacuum experiments, these branching embedded heat pipe networks can efficiently distribute heat over panels for finned surface efficiencies of η f >70% at T H = 500 K input heat. This paper first presents laser powder-bed fusion AM strategies to produce embedded porous structures for wicking heat pipes in Inconel 718 and titanium alloys (commercially pure and Ti-6Al-4V alloys). Post-build chemical and thermal treatments are described that yield hydrophilic wicking surfaces for operation with water. Transient rate-of-rise experiments with water and acetone are reported that yield estimates for AM wick porosity (ϵ), permeability (K), and effective pore radius ( r pore ). Based on the wick characterization results, small prototype radiator panels (75 × 125 mm) with integrated heat pipe networks were manufactured. Heat rejection performance data are presented from cold thermal vacuum testing, with heat input temperatures up to ~510 K. Future efforts will focus on improving heat pipe performance, optimizing radiator mass, and evaluating larger panels to assess scalability.

nuclear electric propulsion↗
NASA NTRS2004

Manufacturing High-Quality Carbon Nanotubes at Lower Cost

A modified electric-arc welding process has been developed for manufacturing high-quality batches of carbon nanotubes at relatively low cost. Unlike in some other processes for making carbon nanotubes, metal catalysts are not used and, consequently, it is not necessary to perform extensive cleaning and purification. Also, unlike some other processes, this process is carried out at atmospheric pressure under a hood instead of in a closed, pressurized chamber; as a result, the present process can be implemented more easily. Although the present welding-based process includes an electric arc, it differs from a prior electric-arc nanotube-production process. The welding equipment used in this process includes an AC/DC welding power source with an integral helium-gas delivery system and circulating water for cooling an assembly that holds one of the welding electrodes (in this case, the anode). The cathode is a hollow carbon (optionally, graphite) rod having an outside diameter of 2 in. (approximately equal to 5.1 cm) and an inside diameter of 5/8 in. (approximately equal to 1.6 cm). The cathode is partly immersed in a water bath, such that it protrudes about 2 in. (about 5.1 cm) above the surface of the water. The bottom end of the cathode is held underwater by a clamp, to which is connected the grounding cable of the welding power source. The anode is a carbon rod 1/8 in. (approximately equal to 0.3 cm) in diameter. The assembly that holds the anode includes a thumbknob- driven mechanism for controlling the height of the anode. A small hood is placed over the anode to direct a flow of helium downward from the anode to the cathode during the welding process. A bell-shaped exhaust hood collects the helium and other gases from the process. During the process, as the anode is consumed, the height of the anode is adjusted to maintain an anode-to-cathode gap of 1 mm. The arc-welding process is continued until the upper end of the anode has been lowered to a specified height above the surface of the water bath. The process causes carbon nanotubes to form in the lowest 2.5 cm of the anode. It also causes a deposit reminiscent of a sandcastle to form on the cathode. The nanotube-containing material is harvested. The cathode and anode can then be cleaned (or the anode is replaced, if necessary) and the process repeated to produce more nanotubes. Tests have shown that the process results in approximately equal to 50-percent yield of carbon nanotubes (mostly of the single-wall type) of various sizes. Whereas the unit cost of purified single-wall carbon nanotubes produced by other process is about $1,000/g in the year 2000, it has been estimated that for the present process, the corresponding cost would be about $10/g.

Benavides, Jeanette M.↗
DOE OSTI2016

Initial Assessment of X-Ray Computer Tomography image analysis for material defect microstructure

The original development work leading to this report was focused on the non destructive three-dimensional (3-D) characterization of nuclear graphite as a means to better understand the nature of the inherent pore structure. The pore structure of graphite and its evolution under various environmental factors such as irradiation, mechanical stress, and oxidation plays an important role in their observed properties and characteristics. If we are to transition from an empirical understanding of graphite behavior to a truly predictive mechanistic understanding the pore structure must be well characterized and understood. As the pore structure within nuclear graphite is highly interconnected and truly 3-D in nature, 3-D characterization techniques are critical. While 3-D characterization has been an excellent tool for graphite pore characterization, it is applicable to a broad number of materials systems over many length scales. Given the wide range of applications and the highly quantitative nature of the tool, it is quite surprising to discover how few materials researchers understand and how valuable of a tool 3-D image processing and analysis can be. Ultimately, this report is intended to encourage broader use of 3 D image processing and analysis in materials science and engineering applications, more specifically nuclear-related materials applications, by providing interested readers with enough familiarity to explore its vast potential in identifying microstructure changes. To encourage this broader use, the report is divided into two main sections. Section 2 provides an overview of some of the key principals and concepts needed to extract a wide variety of quantitative metrics from a 3-D representation of a material microstructure. The discussion includes a brief overview of segmentation methods, connective components, morphological operations, distance transforms, and skeletonization. Section 3 focuses on the application of concepts from Section 2 to relevant materials at Idaho National Laboratory. In this section, image analysis examples featuring nuclear graphite will be discussed in detail. Additionally, example analyses from Transient Reactor Test Facility low-enriched uranium conversion, Advanced Gas Reactor like compacts, and tristructural isotopic particles are shown to give a broader perspective of the applicability to relevant materials of interest.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2014

Influence of Mixed Mode I-Mode II Loading on Fatigue Delamination Growth Characteristics of a Graphite Epoxy Tape Laminate

Mixed mode I-mode II interlaminar tests were conducted on IM7/8552 tape laminates using the mixed-mode bending test. Three mixed mode ratios, G(sub II)/G(sub T) = 0.2, 0.5, and 0.8, were considered. Tests were performed at all three mixed-mode ratios under quasi-static and cyclic loading conditions, where the former static tests were used to determine initial loading levels for the latter fatigue tests. Fatigue tests at each mixed-mode ratio were performed at four loading levels, Gmax, equal to 0.5G(sub c), 0.4G(sub c), 0.3G(sub c), and 0.2G(sub c), where G(sub c) is the interlaminar fracture toughness of the corresponding mixed-mode ratio at which a test was performed. All fatigue tests were performed using constant-amplitude load control and delamination growth was automatically documented using compliance solutions obtained from the corresponding quasi-static tests. Static fracture toughness data yielded a mixed-mode delamination criterion that exhibited monotonic increase in Gc with mixed-mode ratio, G(sub II)/G(sub T). Fatigue delamination onset parameters varied monotonically with G(sub II)/G(sub T), which was expected based on the fracture toughness data. Analysis of non-normalized data yielded a monotonic change in Paris law exponent with mode ratio. This was not the case when normalized data were analyzed. Fatigue data normalized by the static R-curve were most affected in specimens tested at G(sub II)/G(sub T)=0.2 (this process has little influence on the other data). In this case, the normalized data yielded a higher delamination growth rate compared to the raw data for a given loading level. Overall, fiber bridging appeared to be the dominant mechanism, affecting delamination growth rates in specimens tested at different load levels and differing mixed-mode ratios.

Ratcliffe, James G.↗
DOE OSTI2016

AGC-4 Graphite Preirradiation Data Analysis Report

The Advanced Reactor Technology (ART) Graphite R&D program is conducting an extensive graphite irradiation program to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs. Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of new HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance is required for the physical, mechanical, and thermal properties of each major graphite grade with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change—where the graphite physically changes (i.e., first shrinking and then expanding with increasing neutron dose). This disparity in material volume change can produce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the components longer useful lifetimes within the core. Determining the irradiation creep rates of nuclear grade graphite is critical for determining the useful lifetime of graphite components and is a major component of the Advanced Graphite Creep (AGC) experiment. The AGC experiment is currently underway to determine the in service behavior of these new graphite grades for HTR and molten salt reactor designs. This experiment will examine properties and behavior of nuclear grade graphite over a large spectrum of temperatures, irradiation fluencies and applied stress levels that are expected to cause irradiation creep strains within a HTR graphite component. Irradiation data are provided through the AGC test series, which is comprised of six planned capsules irradiated in the Advanced Test Reactor (ATR) in a large flux trap located at Idaho National Laboratory (INL). Each irradiation capsule consists of over 400 graphite specimens that are characterized before and after irradiation to determine the irradiation induced material properties changes and life-limiting irradiation creep rate for each graphite grade.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Uncertainty Quantification of Calculated Temperatures for the AGR 3/4 Experiment

A series of Advanced Gas Reactor (AGR) irradiation experiments are being conducted within the Advanced Reactor Technology (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaign are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and accident conditions, and support development and validation of fuel performance and fission product transport models and codes (PLN 3636, “Technical Program Plan for INL Advanced Reactor Technologies Technology Development Office/Advanced Gas Reactor Fuel Development and Qualification Program”). The AGR 3/4 test was inserted in the Northeast Flux Trap position in the Advanced Test Reactor (ATR) core at Idaho National Laboratory (INL) in December 2011 and successfully completed irradiation in mid-April 2014, resulting in irradiation of the tristructural isotropic (TRISO) fuel for 369.1 effective full-power days (EFPDs) during approximately 2.4 calendar years. The AGR 3/4 data, including the irradiation data and calculated results, were qualified and stored in the Nuclear Data Management and Analysis System (NDMAS). To support the U.S. TRISO fuel performance assessment and to provide data for validation of fuel performance and fission product transport models and codes, the daily as run thermal analysis has been performed separately on each of twelve AGR 3/4 capsules for the entire irradiation as discussed in ECAR-2807, “AGR 3/4 Daily As Run Thermal Analyses”. The ABAQUS code’s finite element-based thermal model predicts the daily average volume average (VA) fuel temperature (FT), peak FT, and graphite matrix, sleeve, and sink temperature in each capsule. The JMOCUP simulation codes were also created to perform depletion calculations for the AGR 3/4 experiment (ECAR-2753, “JMOCUP As-Run Daily Physics Depletion Calculation for the AGR 3/4 TRISO Particle Experiment in ATR Northeast Flux Trap”). This depletion analysis provides fast fluence and fission heat rate data for all components (fuel compacts, graphite rings, stainless steel retainer, etc.), which are used as inputs for the thermal analysis codes. The graphite temperatures from thermocouples (TCs) in each capsule were used to calibrate these thermal analysis codes. However, given the high rate of TC failure under the harsh irradiation and thermal conditions in the AGR capsules, the thermal analysis results are very useful in aiding TC data qualification, increasing the confidence in delineating failures of the measuring instruments (TCs) from physical mechanisms that may have shifted the system thermal response.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2002

Characteristics of Pool Boiling on Graphite-Copper Composite Surfaces

Nucleate pool boiling performance of different liquids on graphite-copper composite (Gr-Cu) surfaces has been experimentally studied and modeled. Both highly wetting fluids, such as freon-113 and pentane, and a moderately wetting fluid (water) were tested on the Gr-Cu surfaces with different graphite-fiber volume fractions to reveal the enhancement effects of the composite surfaces on the nucleate pool boiling. Results of the experiments show that the graphite-fiber volume fraction has an optimum value. The Gr-Cu composite surface with 25 percent graphite-fiber volume (f=0.25) has a maximum enhancement effect on the nucleate boiling heat transfer comparing to the pure copper surface. For the highly wetting fluid, the nucleate boiling heat transfer is generally enhanced on the Gr- Cu composite surfaces by 3 to 6 times shown. In the low heat flux region, the enhancement is over 6 times, but in the high heat flux region, the enhancement is reduced to about 40%. For the moderately wetting fluid (water), stronger enhancement of nucleate boiling heat transfer is achieved on the composite surface. It shown the experimental results in which one observes the nucleate boiling heat transfer enhancement of 5 to 10 times in the low heat flux region and an enhancement of 3 to 5 times in the high heat flux region. Photographs of bubble departure during the initial stage of nucleate boiling indicate that the bubbles detached from the composite surface are much smaller in diameter than those detached from the pure copper surface. Typical photographs are presented.It shows that the bubbles departed from the composite surface have diameters of only O(0.1) mm, while those departed from the pure copper surface have diameters of O(1) mm. It is also found that the bubbles depart from the composite surface at a much higher frequency, thus forming vapor columns. These two phenomena combined with high thermal conductivity of the graphite fiber are considered the mechanisms for such a significant augmentation in nucleate boiling heat transfer on the composite surfaces. A physical model is developed to describe the phenomenon of bubble departure from the composite surface: The preferred site of bubble nucleation is the fiber tip because of higher tip temperature than the surrounding copper base and poor wettability of the graphite tip compared with that of the base material (copper). The high evaporation rate near the contact line produces the vapor cutback due to the vapor recoil pushing the three-phase line outwards from the fiber tip, and so a neck of the bubble is formed near the bubble bottom. Evaporation and surface tension accelerate the necking process and finally result in the bubble departure while a new small bubble is formed at the tip when the surface tension pushes the three-phase line back to the tip. The process is schematically shown. The proposed model is based on and confirmed by experimental results.

Zhang, Nengli↗
DOE OSTI2016

Electrode behavior RE-visited: Monitoring potential windows, capacity loss, and impedance changes in Li 1.03 (Ni 0.5 Co 0.2 Mn 0.3 ) 0.97 O 2 /silicon-graphite full cells

Here, the capacity and power performance of lithium-ion battery cells evolve over time. The mechanisms leading to these changes can often be identified through knowledge of electrode potentials, which contain information about electrochemical processes at the electrode-electrolyte interfaces. In this study we monitor electrode potentials within full cells containing a Li 1.03 (Ni 0.5 Co 0.2 Mn 0.3 ) 0.97 O 2 –based (NCM523) positive electrode, a silicon-graphite negative electrode, and an LiPF6-bearing electrolyte, with and without fluoroethylene carbonate (FEC) or vinylene carbonate (VC) additives. The electrode potentials are monitored with a Li-metal reference electrode (RE) positioned besides the electrode stack; changes in these potentials are used to examine electrode state-of-charge (SOC) shifts, material utilization, and loss of electrochemically active material. Electrode impedances are obtained with a Li x Sn RE located within the stack; the data display the effect of cell voltage and electrode SOC changes on the measured values after formation cycling and after aging. Our measurements confirm the beneficial effect of FEC and VC electrolyte additives in reducing full cell capacity loss and impedance rise after cycling in a 3.0–4.2 V range. Comparisons with data from a full cell containing a graphite-based negative highlight the consequences of including silicon in the electrode. Our observations on electrode potentials, capacity, and impedance changes on cycling are crucial to designing long-lasting, silicon-bearing, lithium-ion cells.

25 ENERGY STORAGE↗