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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 973 records · Page 54
All sourcesNASA NTRSDOE OSTI
NASA NTRS1980

Argon/UF6 plasma experiments: UF6 regeneration and product analysis

An experimental and analytical investigation was conducted to aid in developing some of the technology necessary for designing a self-critical fissioning uranium plasma core reactors (PCR). This technology is applicable to gaseous uranium hexafluoride nuclear-pumped laser systems. The principal equipment used included 1.2 MW RF induction heater, a d.c. plasma torch, a uranium tetrafluoride feeder system, and batch-type fluorine/UF6 regeneration systems. Overall objectives were to continue to develop and test materials and handling techniques suitable for use with high-temperature, high-pressure, gaseous UF6; and to continue development of complementary diagnostic instrumentation and measurement techniques to characterize the effluent exhaust gases and residue deposited on the test chamber and exhaust system components. Specific objectives include: a development of a batch-type UF6 regeneration system employing pure high-temperature fluorine; development of a ruggedized time-of-flight mass spectrometer and associated data acquisition system capable of making on-line concentration measurements of the volatile effluent exhaust gas species in a high RF environment and corrosive environment of UF6 and related halide compounds.

Roman, W. C.↗
NASA NTRS2010

Ares V: Progress Toward Unprecedented Heavy Lift

Ares V represents the vehicle that will again make possible human exploration beyond low Earth orbit. The Ares V is part of NASA s Constellation Program architecture developed to support the International Space Station (ISS), establish a permanent human presence on the Moon, and explore it to an extent far greater than was possible with the Apollo Program. Ares V will carry the lunar lander to orbit where it will join the Orion crew spacecraft, launched by the smaller Ares I launch vehicle. Then the Ares V upper stage will send the Orion and lander to the Moon. Ares V is also intended to launch automated cargo landers to the Moon. The Ares vehicles are designed to employ the proven technologies and experience from the Space Shuttle, Delta IV, and earlier U.S. programs, as well as sharing common components where feasible. The Ares V is in an early stage of concept development. However, commonality allows it to benefit from development work already under way on the Ares I, including the first stage booster, and upper stage, J-2X upper stage engine. This paper will discuss progress to date on the Ares V and its potential for freeing payload designers from current mass and volume constraints. Progress includes development progress on Ares I elements that will be shared by the two launch vehicles. The Ares I first stage recently completed a successful test firing of Development Motor 1 (DM-1). The J-2X engine is proceeding with manufacturing of components for the first development engines that will be used for testing. Several component-level tests have been completed or are under way that will help verify designs and confirm solutions to design challenges. The Ares V Earth departure stage will benefit from the Ares I upper stage development, including design, manufacturing, and materials testing. NASA is also working with government and industry to collect data on flights and testing of the operational RS-68 engine and potential upgrades. The Ares V team continues to evaluate technical options, vehicle configurations, and operations concepts for the Ares V. The team recently completed a Fall Face-to-Face meeting that served as a stepping-stone to the Systems Requirements Review (SRR). This four-day meeting served as an information exchange for the various teams at several NASA field centers and supporting contractors.

Sumrall, Phil↗
DOE OSTI2023

The Innovation of I-Loop

The closing and decommissioning of Halden Boiling Water Reactor in Norway resulted in the loss of many nuclear irradiation testing facilities, notably those providing boiling water reactor (BWR) environments. To lighten the loss of such an important asset, the Advanced Test Reactor (ATR) is being called upon. The recent overhaul of the ATR’s core and the installation of the new Top Head Closure Plate now makes it possible to utilize the ATR’s I-loop locations for I-Loop Tube test facilities. Similar to the current in-pile tube facilities, these new I-Loop facilities will capitalize on the available space and neutron flux the ATR generates in this region of the core for fuels and materials testing. Current efforts are underway to add new PWR and BWR I-Loop Tube facilities to the I-Loop positions for these purposes.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1992

A combined analytical-experimental tensile test technique for brittle materials

A semiconventional tensile test technique is developed for impact ices and other brittle materials. Accurate results have been obtained on ultimate strength and modulus of elasticity in a refrigerated ice test. It is noted that the technique can be used to determine the physical properties of impact ices accreted inside icing wind tunnels or other brittle materials.

Chu, M. L.↗
DOE OSTI2017

Cluster 1: commercializing additive manufacturing—hurdles in materials characterization and testing

A major challenge in the commercialization of additive manufactured (AM) materials and processes is the ability to achieve acceptance of processes and products. There has been some progress towards acceptance has been made by adapting legacy qualification paradigms to match with the very limited process control and monitoring offered by AM machines. The opportunity for in-situ measurement can provide process monitoring and control perhaps changing the way we qualify parts however it is limited by lack of adequate process measurement methods. New measurement techniques, sensors and correlations to relevant phenomena are needed that enable process control and monitoring for consistently producing high quality articles. Beyond process data we need to characterize uncertainties of performance in all aspects of material, process and final part. These are prerequisites to achieving articles that are indeed worthy of materials characterization efforts that establish a microstructural reference of desirable performance through process-structure-property relations. Only then can industry apply physics based understanding of the material, part and process to probabilistically predict performance of an AM part. Our paper provides a brief overview, discussion of hurdles and key areas where R&D investment is needed.

36 MATERIALS SCIENCE↗
NASA NTRS1972

Application of Rochrome liquid crystal tapes for thermographic testing of bonded structures

The use of ROCHROME liquid crystal tape for thermographic testing of bonded structures for aircraft is described. The techniques for applying the tape, heating the test panels, and photographing the resulting thermal patterns are described and discussed. The results obtained are compared with those reported previously for the same test panels coated directly with liquid crystals applied by spraying. Other panels previously examined by various nondestructive techniques, both by thermography, were also tested. Results for these latter panels are presented and discussed. For both sets of panels good correlation with previous work was obtained. Deficiencies associated with present ROCHROME tape are described. In general, these stem from the lack of optimization for materials testing of the ROCHROME tape, particularly with respect to the adhesive system and supporting film.

Source record↗
DOE OSTI2014

A fusion nuclear science facility for a fast-track path to DEMO

An accelerated fusion energy development program, a fast-track approach, requires proceeding with a nuclear and materials testing program in parallel with research on burning plasmas, ITER. A Fusion Nuclear Science Facility (FNSF) would address many of the key issues that need to be addressed prior to DEMO, including breeding tritium and completing the fuel cycle, qualifying nuclear materials for high fluence, developing suitable materials for the plasma-boundary interface, and demonstrating power extraction. The Advanced Tokamak (AT) is a strong candidate for an FNSF as a consequence of its mature physics base, capability to address the key issues, and the direct relevance to an attractive target power plant. The standard aspect ratio provides space for a solenoid, assuring robust plasma current initiation,and for an inboard blanket, assuring robust tritium breeding ratio (TBR) >1 for FNSF tritium self-sufficiency and building of inventory needed to start up DEMO. An example design point gives a moderate sized Cu-coil device with R/a = 2.7 m/0.77 κ = 2.3, B T = 5.4 T, I P = 6.6 MA, β N = 2.75, P fus = 127 MW. The modest bootstrap fraction of f BS = 0.55 provides an opportunity to develop steady state with sufficient current drive for adequate control. Lastly, proceeding with a FNSF in parallel with ITER provides a strong basis to begin construction of DEMO upon the achievement of Q ~ 10 in ITER.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2012

Improved Accelerated Stress Tests Based on Fuel Cell Vehicle Data

UTC will led a top-tier team of industry and national laboratory participants to update and improve DOE’s Accelerated Stress Tests (AST’s) for hydrogen fuel cells. This in-depth investigation will focused on critical fuel cell components (e.g. membrane electrode assemblies - MEA) whose durability represented barriers for widespread commercialization of hydrogen fuel cell technology. UTC had access to MEA materials that had accrued significant load time under real-world conditions in PureMotion® 120 power plant used in transit buses. These materials are referred to as end-of-life (EOL) components in the rest of this document. Advanced characterization techniques were used to evaluate degradation mode progress using these critical cell components extracted from both bus power plants and corresponding materials tested using the DOE AST’s. These techniques were applied to samples at beginning-of-life (BOL) to serve as a baseline. These comparisons advised the progress of the various failure modes that these critical components were subjected to, such as membrane degradation, catalyst support corrosion, platinum group metal dissolution, and others. Gaps in the existing ASTs predicted the degradation observed in the field in terms of these modes were outlined. Using the gaps, new AST’s were recommended and tested to better reflect the degradation modes seen in field operation. Also, BOL components were degraded in a test vehicle at UTC designed to accelerate the bus field operation.

08 HYDROGEN↗
DOE OSTI2018

Addressing Research and Development Gaps for Plasma-Material Interactions with Linear Plasma Devices

Plasma-material interactions in future fusion reactors have been identified as a knowledge gap to be dealt with before any next step device past ITER can be built. The challenges are manifold. They are related to power dissipation so that the heat fluxes to the plasma-facing components can be kept at technologically feasible levels; maximization of the lifetime of divertor plasma-facing components that allow for steadystate operation in a reactor to reach the neutron fluence required; the tritium inventory (storage) in the plasma-facing components, which can lead to potential safety concerns and reduction in the fuel efficiency; and it is related to the technology of the plasma-facing components itself, which should demonstrate structural integrity under the high temperatures and high neutron fluence. While the dissipation of power exhaust can and should be addressed in high power toroidal devices, the interaction of the plasma with the materials can be best addressed in dedicated linear devices due to their cost effectiveness and ability to address urgent research and development gaps more timely. However, new linear plasma devices are needed to investigate the PMI under fusion reactor conditions and test novel plasma-facing components. Existing linear devices are limited either in their flux, their reactor-relevant plasma transport regimes in front of the target, their fluence, or their ability to test material samples a priori exposed to high neutron fluence. The proposed Material Plasma Exposure eXperiment (MPEX) is meant to address those deficiencies and will be designed to fulfill the fusion reactor-relevant plasma parameters as well as the ability to expose a priori neutron activated materials to plasmas.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Bacterial Membranes Are More Perturbed by the Asymmetric Versus Symmetric Loading of Amphiphilic Molecules

Characterizing the biophysical properties of bacterial membranes is critical for understanding the protective nature of the microbial envelope, interaction of biological membranes with exogenous materials, and designing new antibacterial agents. Presented here are molecular dynamics simulations for two cationic quaternary ammonium compounds, and the anionic and nonionic form of a fatty acid molecule interacting with a Staphylococcus aureus bacterial inner membrane. The effect of the tested materials on the properties of the model membranes are evaluated with respect to various structural properties such as the lateral pressure profile, lipid tail order parameter, and the bilayer’s electrostatic potential. Conducting asymmetric loading of molecules in only one leaflet, it was observed that anionic and cationic amphiphiles have a large impact on the Staphylococcus aureus membrane’s electrostatic potential and lateral pressure profile as compared to a symmetric distribution. Nonintuitively, we find that the cationic and anionic molecules induce a similar change in the electrostatic potential, which points to the complexity of membrane interfaces, and how asymmetry can induce biophysical consequences. Finally, we link changes in membrane structure to the rate of electroporation for the membranes, and again find a crucial impact of introducing asymmetry to the system. Understanding these physical mechanisms provides critical insights and viable pathways for the rational design of membrane-active molecules, where controlling the localization is key.

59 BASIC BIOLOGICAL SCIENCES↗
NASA NTRS1998

Crack Branching and Fracture Mirror Data of Glasses and Advanced Ceramics

The fracture mirror and crack branching constants were determined from three glasses and nine advanced ceramics tested under various loading and specimen configurations in an attempt to use the constants as a data base for fractography. The ratios of fracture mirror or crack branching constant to fracture toughness were found to be approximately two for most ceramic materials tested. A demonstration of how to use the two constants as a tool for verifying stress measurements was presented for silicon nitride disk specimens subjected to high-temperature, constant stress-rate biaxial flexure testing.

Choi, Sung R.↗
DOE OSTI2017

United States Advanced Ultra-Supercritical Component Test Facility for 760°C Steam Power Plants ComTest Project

Following the successful completion of a 15-year effort to develop and test materials that would allow coal-fired power plants to be operated at advanced ultra-supercritical (A-USC) steam conditions, a United States-based consortium is presently engaged in a project to build an A-USC component test facility (ComTest). A-USC steam cycles have the potential to improve cycle efficiency, reduce fuel costs, and reduce greenhouse gas emissions. Current development and demonstration efforts are focused on enabling the construction of A-USC plants, operating with steam temperatures as high as 1400°F (760°C) and steam pressures up to 5000 psi (35 MPa), which can potentially increase cycle efficiencies to 47% HHV (higher heating value), or approximately 50% LHV (lower heating value), and reduce CO 2 emissions by roughly 25%, compared to today’s U.S. fleet. A-USC technology provides a lower-cost method to reduce CO 2 emissions, compared to CO 2 capture technologies, while retaining a viable coal option for owners of coal generation assets. Among the goals of the ComTest facility are to validate that components made from advanced nickel-based alloys can operate and perform under A-USC conditions, to accelerate the development of a U.S.-based supply chain for the full complement of A-USC components, and to decrease the uncertainty of cost estimates for future A-USC power plants. The configuration of the ComTest facility would include the key A-USC technology components that were identified for expanded operational testing, including a gas-fired superheater, high-temperature steam piping, steam turbine valve, and cycling header component. Membrane walls in the superheater have been designed to operate at the full temperatures expected in a commercial A-USC boiler, but at a lower (intermediate) operating pressure. This superheater has been designed to increase the temperature of the steam supplied by the host utility boiler up to 1400°F (760°C). The steam turbine stop and control valve component has been designed to operate at full A-USC temperatures, and would be tested both in throttling operation and to accumulate accelerated, repetitive stroke cycles. A cycling header component has been designed to confirm the suitability of new high-temperature nickel alloys to cycling operation, expected of future coal-fired power plants. Current test plans would subject these components to A-USC operating conditions for at least 8,000 hours by September 2020. The ComTest project is managed by Energy Industries of Ohio, and technically directed by the Electric Power Research Institute, Inc., with General Electric designing the A-USC components. This consortium is completing the Detailed Engineering phase of the project, with procurement scheduled to begin in late 2017. The effort is primarily funded by the U.S. Department of Energy, through the National Energy Technology Laboratory, along with the Ohio Development Services Agency. This presentation outlines the motivation for the project, explains the project’s structure and schedule, and provides technical details on the design of the ComTest facility.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2014

CORROSION OF ALUMINUM CLAD SPENT NUCLEAR FUEL IN THE 70 TON CASK DURING TRANSFER FROM L AREA TO H-CANYON

Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 260 {degrees}C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 {degrees}C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Corrosion of Aluminum Clad Spent Nuclear Fuel in the 70 Ton Cask during Transfer from L Area to H-Canyon

Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Corrosion of Aluminum Clad Spent Nuclear Fuel in the 70 Ton Cask during Transfer from L Area to H-Canyon (Rev.1)

Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33% was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Joint Workplan on Filler Investigations for DPCs.

This workplan addresses filler attributes (i.e., possible requirements), assumptions needed for analysis, selection of filler materials, testing needs, and a long-range perspective on R&D activities leading to filler demonstration and a safety basis for implementation.

36 MATERIALS SCIENCE↗
NASA NTRS1991

Ion beam textured and coated surfaces experiment (IBEX)

The IBEX, with 36 samples of various materials, was placed aboard the LDEF. Twenty-seven of the samples had surfaces modified using ion beam technology, and nine were made up of commercially available materials. The materials are in some way useful in space power systems. The various types of materials tested included six categories: (1) ion beam structured surfaces suitable for solar thermal (concentrator) or space radiators; (2) ion beam sputtered conductive coatings for thermal and space charge control; (3) solar reflector surfaces; (4) flexible thin film coatings and solar array blanket material for protection of spacecraft polymers; (5) painted and/or state-of-the-art solar thermal materials; and (6) micrometeoroid sensitive detector. Data analysis presented include the optical properties of each surface before and after exposure to the space environment and the respective backup surfaces.

Mirtich, Michael J.↗
NASA NTRS1991

Surface fatigue life of M50NiL and AISI 9310 spur gears and R C bars

Spur gear endurance tests and rolling element surface fatigue tests were conducted to study vacuum induction melted, vacuum arc remelted (VIM-VAR) M50NiL steel for use as a gear steel in advanced aircraft applications, to determine its endurance characteristics, and to compare the results with those for standard VAR and VIM-VAR AISI 9310 gear material. Tests were conducted with spur gears and rolling contact bars manufactured from VIM-VAR M50NiL and VAR and VIM-VAR AISI 9310. The gear pitch diameter was 8.9 cm. Gear test conditions were an inlet oil temperature of 320 K, and outlet oil temperature of 350 K, a maximum Hertz stress of 1.71 GPa, and a speed of 10000 rpm. Bench rolling element fatigue tests were conducted at ambient temperatures with a bar speed of 12,500 rpm and a maximum Hertz stress of 4.83 GPa. The VIM-VAR M50NiL gears had a surface fatigue life that was 4.5 and 11.5 times that for VIM-VAR and VAR AISI 9310 gears, respectively. The surface fatigue life of the VIM-VAR M50NiL rolling contact bars was 13.2 and 21.6 times that for the VIM-VAR and VAR AISI 9310, respectively. The VIM-VAR M50NiL material was shown to have good resistance to fracture through a fatigue spall and superior fatigue life to both other gears.

Townsend, Dennis P.↗