Fabrication process developments of a structural model for research on hydrogen- fueled hypersonic vehicles
Test structural model for hydrogen-fueled hypersonic vehicles
TEST EQUIPMENT↗
SEARCH · Search NASA
Results for “Fuel testing”
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.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
Assessment of Low Enriched Uranium Fueled Core Configurations for the Versatile Test Reactor
The Versatile Test Reactor (VTR) Program was initiated in FY-2017. This multi-laboratory effort supported by the U.S. Department of Energy has the long-term objective of completing a Versatile Test Reactor concept with associated cost and schedule estimate that enables a fully informed DOE decision within 3 years, to have a fully operating VTR by 2026 if acquisition is authorized, and to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The VTR concept uses sodium as coolant, features a core power of 300 MWth, and employs reactor-grade plutonium as fissile material in a U-Pu-10Zr fuel alloy. The reference core layout provides 10 assembly locations for high-flux irradiation testing with peak fast flux in excess of 4x1015 n/cm2-s. In order to establish a comparison point with other types of fuel, alternative core configurations were investigated using Low Enriched Uranium (LEU) fuel. The objective of this study was to modify the reference VTR core such as to use LEU fuel and assess the performance tradeoff, while retaining the same core envelope.
Stauff, Nicolas E.↗
Lunar Regenerative Fuel Cell (RFC) Reliability Testing for Assured Mission Success
NASA's Constellation program has selected the closed cycle hydrogen oxygen Polymer Electrolyte Membrane (PEM) Regenerative Fuel Cell (RFC) as its baseline solar energy storage system for the lunar outpost and manned rover vehicles. Since the outpost and manned rovers are "human-rated," these energy storage systems will have to be of proven reliability exceeding 99 percent over the length of the mission. Because of the low (TRL=5) development state of the closed cycle hydrogen oxygen PEM RFC at present, and because there is no equivalent technology base in the commercial sector from which to draw or infer reliability information from, NASA will have to spend significant resources developing this technology from TRL 5 to TRL 9, and will have to embark upon an ambitious reliability development program to make this technology ready for a manned mission. Because NASA would be the first user of this new technology, NASA will likely have to bear all the costs associated with its development.When well-known reliability estimation techniques are applied to the hydrogen oxygen RFC to determine the amount of testing that will be required to assure RFC unit reliability over life of the mission, the analysis indicates the reliability testing phase by itself will take at least 2 yr, and could take up to 6 yr depending on the number of QA units that are built and tested and the individual unit reliability that is desired. The cost and schedule impacts of reliability development need to be considered in NASA's Exploration Technology Development Program (ETDP) plans, since life cycle testing to build meaningful reliability data is the only way to assure "return to the moon, this time to stay, then on to Mars" mission success.
Bents, David J.↗
Active Control of High Frequency Combustion Instability in Aircraft Gas-Turbine Engines
Active control of high-frequency (greater than 500 Hz) combustion instability has been demonstrated in the NASA single-nozzle combustor rig at United Technologies Research Center. The combustor rig emulates an actual engine instability and has many of the complexities of a real engine combustor (i.e. actual fuel nozzle and swirler, dilution cooling, etc.) In order to demonstrate control, a high-frequency fuel valve capable of modulating the fuel flow at up to 1kHz was developed. Characterization of the fuel delivery system was accomplished in a custom dynamic flow rig developed for that purpose. Two instability control methods, one model-based and one based on adaptive phase-shifting, were developed and evaluated against reduced order models and a Sectored-1-dimensional model of the combustor rig. Open-loop fuel modulation testing in the rig demonstrated sufficient fuel modulation authority to proceed with closed-loop testing. During closed-loop testing, both control methods were able to identify the instability from the background noise and were shown to reduce the pressure oscillations at the instability frequency by 30%. This is the first known successful demonstration of high-frequency combustion instability suppression in a realistic aero-engine environment. Future plans are to carry these technologies forward to demonstration on an advanced low-emission combustor.
Corrigan, Bob↗
ELECTRON PROBE MICROANALYSIS OF IRRADIATED AND 1600°C SAFETY-TESTED AGR-1 TRISO FUEL PARTICLES WITH LOW AND HIGH RETAINED 110MAG
AGR-1 fuel Compact 4-3-3 achieved 18.63% FIMA and was exposed subsequently to a safety test at 1600°C. Two particles, AGR1-433-003 and AGR1-433-007, with measured-to-calculated 110mAg inventories of <22% and 100%, respectively, were selected for comparative electron microprobe analysis to determine whether the distribution or abundance of fission products differed proximally and distally from the deformed kernel in AGR1-433-003, and how this compared to fission product distribution in AGR1-433-007. On the deformed side of AGR1-433-003, Xe, Cs, I, Eu, Sr, and Te concentrations in the kernel buffer interface near the protruded kernel were up to six times higher than on the opposite, non-deformed side. At the SiC-inner pyrolytic carbon (IPyC) interface proximal to the deformed kernel, Pd and Ag concentrations were 1.2 wt% and 0.04 wt% respectively, whereas on the SiC-IPyC interface distal from the kernel deformation those elements measured 0.4 and 0.01 wt%, respectively. Palladium and Ag concentrations at the SiC-IPyC interface of AGR1-433-007 were 2.05 and 0.05 wt.%, respectively. Rare earth element concentrations at the SiC-IPyC interface of AGR1-433-007 were a factor of ten higher than at the SiC-IPyC interfaces measured in particle AGR1-433-003. Palladium permeated the SiC layer of AGR1-433-007 and the non-deformed SiC layer of AGR1-433-003.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Microstructure and fission products in the UCO kernel of an AGR-1 TRISO fuel particle after post irradiation safety testing
Tristructural isotropic (TRISO) coated particle fuel as a key fuel concept for the High Temperature Gas-cooled Reactors (HTGRs) and a candidate accident tolerant fuel (ATF) has been currently investigated under the US-DOE Advanced Gas Reactor (AGR) fuel development and qualification program. Over the past decades, extensive studies have been conducted to evaluate the fission product release, diffusion on Ag, Pd and Cd in the SiC layer and the TRISO coating system, and post-safety-test performance. However, to date, there are limited reported results on the fuel kernels’ response to irradiation with or without post irradiation high temperature annealing. To incrementally fulfill this knowledge gap extensive studies using transmission electron microscopes (TEM) and atom probe tomography (APT) were conducted on a TRISO fuel particle kernel with a 19.74% 235U enrichment, irradiated to 18.63% FIMA and subsequently subjected to post-safety testing at 1600ºC for 300 hours.. Microstructure characterizations, elemental analysis, and phase identification were conducted using conventional TEM and scanning TEM (STEM) imaging, energy-dispersive X-ray spectroscopy (EDS), selected-area electron diffraction (SAED) and atom probe tomography (APT). The following findings were made: (1) significant reconstructions and phase evolutions occurred in the irradiated and post-safety annealed fuel kernel, and its microstructure consists of two primary phases, namely a “high-Z (atomic mass)” UC phase and a “low-Z” UO phase, (2) no fission gas bubbles are identified within the fuel kernel, that can be attributed to the high temperature post safety annealing, (3) fission products Zr, Nb, Mo, Ru, Tc and Rh were found to preferentially segregate into UC phase or to form metallic precipitates, while the lanthanide fission products tend to stay in the solution of UO phase, (4) Pd was detected in the rod-shaped precipitates, (5) APT data reveals segregation of Te, Pd, and Ag at the UO and UC phase boundary, and (6) Pd is detected in both the UO and UC phases but with an enrichment in the former one.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Testing Instrument Extremes in the TREAT Facility
TREAT is a transient power-shaping reactor whose primary mission is to enhance safety performance by testing nuclear fuels and materials under thermodynamic and neutronic conditions ranging from off-normal to extreme. Transient testing of nuclear fuels is analogous to car crash testing, in that the dramatic changes commonly seen when comparing the test specimens' initial and final state points necessitate in-situ measuring in order to interpret and understand the evolution of the experiment. TREAT’s experiment design strategy utilizes a highly reconfigurable and accessible reactor core that affords flexibility when installing experiment devices, thanks to the lack of a primary coolant boundary and the inclusion of multiple access points in the reactor bioshield. Within the experiment devices, the test specimens (which are integrated with their test environments), desired instrumentation diagnostics, and safety considerations are all contained in a single engineered package. The experiments encompass various types of environments (e.g., water, sodium, and gas at temperature and pressure), as well as a range of control options (e.g., static or flowing coolants). The experimental conditions are uniquely representative of advanced reactors, providing challenges to sensor performance but also affording a unique opportunity for development and qualification of sensors suited to these environments.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
HYPULSE combustor analysis
The analysis of selected data from tests of unit fuel injectors in a generic scramjet combustor model is presented. The tests were conducted in the NASA HYPULSE expansion tube at conditions typical of flight at Mach 13.5 and 17. The analysis used a three-stream tube method, with finite-rate chemistry, in which the fuel, test gas, and mixing/combustive streams were treated independently but with the same static pressure. Performance of three candidate fuel injectors is examined based on deduced mixing and combustion efficiencies.
Rizkalla, O. F.↗
PIE on Safety-Tested Loose Particles from Irradiated Compact 4-4-2
Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High Temperature Gas-cooled Reactors (HTGRs). This work is sponsored by the Department of Energy Office of Nuclear Energy (DOE-NE) through the Advanced Reactor Technologies (ART) Office under the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program. The AGR-1 experiment was the first in a series of TRISO fuel irradiation tests initiated in 2006. The AGR-1 TRISO particles and fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 using laboratory-scale equipment and irradiated for 3 years in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. Post-irradiation examination was performed at INL and ORNL to study how the fuel behaved during irradiation, and to test fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing and post-safety testing PIE conducted at ORNL on loose particles extracted from irradiated AGR-1 Compact 4-4-2.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Hot Hydrogen Testing of Tungsten-Uranium Dioxide (W-UO2) CERMET Fuel Materials for Nuclear Thermal Propulsion
CERMET fuel materials are being developed at the NASA Marshall Space Flight Center for a Nuclear Cryogenic Propulsion Stage. Recent work has resulted in the development and demonstration of a Compact Fuel Element Environmental Test (CFEET) System that is capable of subjecting depleted uranium fuel material samples to hot hydrogen. A critical obstacle to the development of an NCPS engine is the high-cost and safety concerns associated with developmental testing in nuclear environments. The purpose of this testing capability is to enable low-cost screening of candidate materials, fabrication processes, and further validation of concepts. The CERMET samples consist of depleted uranium dioxide (UO2) fuel particles in a tungsten metal matrix, which has been demonstrated on previous programs to provide improved performance and retention of fission products1. Numerous past programs have utilized hot hydrogen furnace testing to develop and evaluate fuel materials. The testing provides a reasonable simulation of temperature and thermal stress effects in a flowing hydrogen environment. Though no information is gained about radiation damage, the furnace testing is extremely valuable for development and verification of fuel element materials and processes. The current work includes testing of subscale W-UO2 slugs to evaluate fuel loss and stability. The materials are then fabricated into samples with seven cooling channels to test a more representative section of a fuel element. Several iterations of testing are being performed to evaluate fuel mass loss impacts from density, microstructure, fuel particle size and shape, chemistry, claddings, particle coatings, and stabilizers. The fuel materials and forms being evaluated on this effort have all been demonstrated to control fuel migration and loss. The objective is to verify performance improvements of the various materials and process options prior to expensive full scale fabrication and testing. Post test analysis will include weight percent fuel loss, microscopy, dimensional tolerance, and fuel stability.
Hickman, Robert↗
The response of accident tolerant fuel cladding to LOCA burst testing: A comparative study of leading concepts
Accident tolerant fuel claddings seek to improve safety margins during loss-of-coolant accident (LOCA) scenarios by reducing cladding steam oxidation rates, hydrogen production, delaying or mitigating cladding burst, and reducing cladding deformation. After rapid and extensive progress in recent years, Cr coated Zr-based alloys, ferritic FeCrAl, and SiC/SiC ceramic matrix composites (CMCs) have emerged as the most promising cladding concepts, but there are few studies comparing the accident tolerance of all three concepts simultaneously. Here, in this study, LOCA burst testing has been conducted on Cr high impulse power magnetron sputtered Zry-4 (Cr/Zry-4), C26M (FeCrAl), and SiC/SiC SiGA (CMC). Experimental observations indicate Cr/Zry-4 burst at higher temperatures than Zry-4 however the effectiveness of the coating on mitigating ZrO 2 formation via Cr 2 O 3 formation depended on physical proximity to the burst opening. No dramatic improvement in diametrical strain nor opening geometry was observed. C26M burst at higher temperatures than Zry-4 at lower pressures, exhibiting small amounts of strain and smaller openings, and formed slow-growing Al 2 O 3 regardless of location. The CMCs did not burst or show any macroscopic signs of deformation, all the way up to an internal cladding pressure of 15.2 MPa. High temperature expanding Nb plug testing indicated that CMC room temperature strength may persist up to 1650°C, with an approximately 40% loss of strength at 1900°C. CMC steam oxidation was also compared to reference chemically vapor deposited SiC up to 1700°C, and CMCs were found to oxidize at slightly accelerated rates that were still orders of magnitude lower than bare Zry-4. Projected H 2 generation during LOCA burst testing showed that when accounting for cladding burst exposing inner diameters to steam, Cr coated Zry-4 reduced H 2 generation by a factor of two compared with bare Zry-4, while both C26M and SiC/SiC reduced H 2 generation by orders of magnitude relative to bare Zry-4.
36 MATERIALS SCIENCE↗
Accident tolerant, non-oxidizable MOX fuel fabrication
Thorium dioxide (ThO{sub 2}) has a set of physical-chemical properties that make it well suited to serve as a nuclear fuel matrix: it has a very high melting point, it cannot be further oxidized, it has low leachability in water, reasonable thermal conductivity (better than UO{sub 2}), and of course it is fertile. Collectively, these properties allow for a fuel ceramic with increased in / out of core stability. As such, Thor Energy and a consortium of entities that include Westinghouse, KAERI, Fortum, NNL, ITU and IFE have recently fabricated (Th,Pu)O{sub 2} pellet specimens in Norway. The Th-MOX production methodology was developed over the course of 2 years of ceramic production trials involving Ce-Th oxide mixtures. It was understood that parameters developed for CeO{sub 2} material might need to have been modified if trial results with PuO{sub 2} led to sub-optimal pellets. Among observations made in this work, the most important production parameter was the preparation of constituent / mixed powders. Correct powder morphology was key to achieving high fuel densities. The process parameters developed for (Th,Ce)O{sub 2} pellets were adopted with the expectation that these would serve as a sound basis for obtaining dense, homogeneous (Th,Pu)O{sub 2} fuel ceramic. The Th-MOX fuel manufacture campaign was successful in producing enough (Th,Pu)O{sub 2} pellets to assemble two 30 cm long fuel pins for testing in the Halden test reactor. These fuel rods will be fitted with various instruments to enable temperature, fission gas release and dimensional changes to be measured while operating at power in conditions similar to those of a BWR.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Research and development of high thermal stability fuels
Increases in aircraft performance are leading to increases in the thermal stress on the primary aircraft coolant--the fuel. Fuel thermal stability limitations may offset future aircraft performance gains. The Air Force's Wright Laboratory is sponsoring several research programs to address this problem. The development of an additive package for JP-8 to improve its thermal stability is the primary focus of this paper. This program involves extensive testing of fuels and additives in a variety of test devices, culminating in tests in a fuel system simulator and engine tests. These tests involve Air Force personnel, on-site contractors (University of Dayton Research Institute, Systems Research Laboratories), Pratt and Whitney Aircraft Co., additive manufacturers, and Sandia National Laboratory. The test devices include several flowing and static tests, where the behavior of a fuel is investigated in a wide variety of environments. The study of several baseline fuels in these devices has led to some new insights into the mechanisms of fuel thermal (in)stability. It is becoming clear that a fuel's tendency to oxidize (to form peroxides, for example) is often inversely proportional to its tendency to form insoluble deposits.
T Edwards↗
Teledyne Energy Systems, Inc., Proton Exchange Member (PEM) Fuel Cell Engineering Model Powerplant. Test Report: Initial Benchmark Tests in the Original Orientation
Proton Exchange Membrane (PEM) fuel cell technology is the leading candidate to replace the alkaline fuel cell technology, currently used on the Shuttle, for future space missions. During a 5-yr development program, a PEM fuel cell powerplant was developed. This report details the initial performance evaluation test results of the powerplant.
Loyselle, Patricia↗
PtCo Cathode Catalyst Morphological and Compositional Changes after PEM Fuel Cell Accelerated Stress Testing
Development of Pt catalysts alloyed with transition metals has led to a new class of state-of-the-art electrocatalysts for oxygen reduction at the cathode of proton exchange membrane fuel cells; however, the durability of Pt-based alloy catalysts is challenged by poor structural and chemical stability. There is a need for better understanding of the morphological and compositional changes that occur to the catalyst under fuel cell operation. In this work, we report in-depth characterization results of a Pt-Co electrocatalyst incorporated in the cathode of membrane electrode assemblies, which were evaluated before and after accelerated stress tests designed specifically to enhance catalyst degradation. Electron microscopy, spectroscopy, and 3D electron tomography analyses of the Pt-Co nanoparticle structures suggest that the small- and intermediate-sized Pt-Co particles, which are typically Pt-rich in the fresh condition, undergo minimal morphological changes, whereas intermediate- and larger-sized Pt-Co nanoparticles that exhibit a porous “spongy” morphology and initially have a higher Co content, transform into hollowed-out shells, which is driven by continuous leaching of Co from the Pt-Co catalysts. We further show how these primary Pt-Co nanoparticle morphologies group toward a lower Co, larger size portion of the size vs. composition distribution, and provide details of their nanoscale morphological features.
30 DIRECT ENERGY CONVERSION↗
Peak-Seeking Control For Reduced Fuel Consumption: Flight-Test Results For The Full-Scale Advanced Systems Testbed FA-18 Airplane
A peak-seeking control algorithm for real-time trim optimization for reduced fuel consumption has been developed by researchers at the National Aeronautics and Space Administration (NASA) Dryden Flight Research Center to address the goals of the NASA Environmentally Responsible Aviation project to reduce fuel burn and emissions. The peak-seeking control algorithm is based on a steepest-descent algorithm using a time-varying Kalman filter to estimate the gradient of a performance function of fuel flow versus control surface positions. In real-time operation, deflections of symmetric ailerons, trailing-edge flaps, and leading-edge flaps of an F/A-18 airplane are used for optimization of fuel flow. Results from six research flights are presented herein. The optimization algorithm found a trim configuration that required approximately 3 percent less fuel flow than the baseline trim at the same flight condition. This presentation also focuses on the design of the flight experiment and the practical challenges of conducting the experiment.
flight optimization↗
Reactivation of the Transient Reactor Test (TREAT) Facility Neutron Radiography Program
Reactivation of the Transient Reactor Test (TREAT) Facility Neutron Radiography Program The Transient Reactor Test (TREAT) Facility is a 19,000 MW reactor located at the Idaho National Laboratory designed to perform transient testing on nuclear fuel experiments and materials. A neutron radiography facility was installed with the first radiograph being produced in 1967 that utilized the indirect (or transfer) method of neutron radiography to examine irradiated nuclear fuels and experiment test loops. The TREAT neutron radiography facility performed approximately 5,000 radiographs by the spring of 1977. The original radiography facility was upgraded in 1974-1975 to correct previous deficiencies and to improve capabilities. In 1986, an upgraded aperture was installed to increase the resolution of images followed by a major reactor upgrade completed in 1988. Overall, the existing neutron radiography system was in good working order. Originally built in 1958, the TREAT Facility was in operation until it was placed in a shutdown status in 1994. Following the Fukishima disaster and seeing a need for enhanced accident tolerant fuels, the United Stated Department of Energy decided to restart the TREAT facility and resume transient operations. In November 2017, the TREAT reactor was successfully restarted and is currently performing operational testing in preparation for initial experiment irradiations and transient testing. This paper reports on efforts to reactivate the TREAT neutron radiography station. A few minor shielding and system improvement upgrades have been completed along with the installation of a dedicated room for computed radiography equipment. Detailed beam characterization work was not performed following the last reactor upgrade. To characterize the neutron beam, gold foil activation measurements were made to determine an average neutron flux and flux profile. Additional multi-foil activation measurements will be made to calculate the energy spectrum. An open beam image will be taken that will provide the information about variations in the beam profile. A series of system qualification shots will also take place to accurately determine the effective image quality and image acquisition parameters. Full operational status of the TREAT neutron radiography system is scheduled for late summer of 2018.
36 MATERIALS SCIENCE↗
Seismic and LOCA safety testing and analysis program for Westinghouse PWR fuel under end of life conditions
The NRC issued Information Notice - IN 2012-09 in June 2012 which challenges approved assumptions based on the beginning of life (BOL) conditions in seismic and LOCA analysis. Westinghouse manufactures a variety of PWR fuel designs, including 14*14, 15*15, 16*16 and 17*17 array, also different fuel lengths from 12 feet to 14 feet. The fuels are supplied to the cores designed by Westinghouse and Combustion Engineering, and other cores around world. To address IN 2012-09 efficiently and economically, Westinghouse and PWR Owner Groups launched a testing and analysis program funded by PWROG. The program planed testing and analysis matrices based on types of fuel designs. The testing plan includes grid crush tests, fuel assembly mechanical tests and flowing water damping tests. Applying modelling technique, the program will reduce the number of test and cost. The ultimate goal of the program is to generate fuel assembly EOL models for all fuel designs manufactured by Westinghouse and to be ready for any plant specific seismic and LOCA analysis to address NRC IN for EOL effect. The phase I of PWROG for EOL conditions has been completed and several conclusions are obtained. First, the gapped grids reduce grid dynamic stiffness, local impact stiffness and crush strength. Also, due to gapped grids, both the fuel assembly natural frequency and stiffness are reduced. Secondly, applying higher fuel assembly damping due to the flowing water effect can significantly reduce fuel assembly impact forces. However, during a seismic event the reactor coolant pumps may trip resulting in a coast-down. As such, reduced flow rates that occur during the pump coast-down must be considered. Thirdly, the analytical model representation of a reactor core adequately characterizes the core response under seismic and LOCA excitations. Static and dynamic structural test results of a fuel assembly compared with analytical predictions indicate that the fuel assembly lateral and axial structural behavior can be fully determined with finite element models. Fourthly, analyses for EOL fuel assemblies under seismic and LOCA loads demonstrate EOL effect and flowing water damping effect. With flowing water damping benefit, significant margin can be recovered for fuel assemblies with EOL conditions and under seismic loads.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗