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At least 1,297 records · Page 72
All sourcesNASA NTRSDOE OSTI
NASA NTRS2000

Oxygen Compatibility Testing of Composite Materials

The development of polymer composite liquid oxygen LO2 tanks is a critical step in creating the next generation of launch vehicles. Future launch vehicles need to minimize the gross liftoff weight (GLOW), which is possible due to the 25%-40% reduction in weight that composite materials could provide over current aluminum technology. Although a composite LO2 tank makes these weight savings feasible, composite materials have not historically been viewed as "LO2 compatible." To be considered LO2 compatible, materials must be selected that will resist any type of detrimental, combustible reaction when exposed to usage environments. This is traditionally evaluated using a standard set of tests. However, materials that do not pass the standard tests can be shown to be safe for a particular application. This paper documents the approach and results of a joint NASA/Lockheed Martin program to select and verify LO2 compatible composite materials for liquid oxygen fuel tanks. The test approach developed included tests such as mechanical impact, particle impact, puncture, electrostatic discharge, friction, and pyrotechnic shock. These tests showed that composite liquid oxygen tanks are indeed feasible for future launch vehicles.

Graf, Neil A.↗
NASA NTRS2020

Analysis of a 100-W Regenerative Fuel Cell Demonstration

A 100-W regenerative fuel cell (RFC) demonstration was successfully completed at the NASA Glenn Research Center on December 11, 2014. This effort was funded by Advanced Space Power Systems milestone and represented the first demonstration of passive laboratory-scale stacks arranged as an RFC, consisting of a non-flow-through fuel cell and a static-feed electrolyzer. Once developed, these stacks present the potential to minimize system mass, volume, and parasitic power in comparison to existing terrestrial electrochemical systems and nonnuclear power generation options. Testing involved the fuel cell powering a simulated load then recycling fuel cell product water by subsequently electrolyzing it to regenerate hydrogen and oxygen gases for fuel cell operations. A total of five cycles were completed over a 2-week period, yielding cell voltage data for stack performance and quality measures for regenerated reactants. Overall, round-trip efficiency was calculated to be 47.4 percent. Gas crossover was observed in electrolyzer gas products, as residual gas analysis showed 1.1 percent hydrogen in oxygen and 0.4 percent oxygen in hydrogen. Fuel cell product water was determined to be more acidic and conductive and have a higher fluoride content when compared to deionized water.

fuel cell↗
DOE OSTI2016

Simulation methodology for fuel assembly drop accident during handling

Fresh and irradiated fuel assembly handling is a routine part of Nuclear Power Plant activities. Fuel Assembly handling occurs during fresh fuel receipt, core refueling, spent fuel assemblies transfer from spent fuel pool to the storage containers, etc. Westinghouse, ENUSA and KEPCO-NF have collaborated in the technical area of Fuel Assembly Drop Accident (FADA) Simulation to update the approach for estimating the number of failed fuel rods as a consequence of a fresh or irradiated fuel assembly drop, which might occur during a handling accident in a PWR-type reactor. The proposed FADA simulation methodology starts with an assessment of fuel assembly impact velocity at the final stage of the drop in water or air. The vertical, horizontal or inclined drop directions are considered. This velocity is an input to determine a fuel assembly structural response during impact which defines the loading function for the fuel assembly components including spacer grids. The dynamic load on the fuel rod is calculated from the detailed spacer grid model and the corresponding loading function. Finally, the load from the spacer grid spring and dimples are applied to the fuel rod model to determine the cladding parameters for use in compliance to fuel rod cladding failure criteria. A combination of first principal, finite element analysis (FEA) and computational fluid dynamic (CFD) methods has been utilized to create a technically sound methodology. A number of specific and traditional tests for the fuel assembly and its components has been used to confirm that the proposed approaches which describe the physical phenomena related to the fuel assembly drop accident are reasonably conservative.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1977

NASA/Pratt and Whitney experimental clean combustor program: Engine test results

A two-stage vorbix (vortex burning and mixing) combustor and associated fuel system components were successfully tested in an experimental JT9D engine at steady-state and transient operating conditions, using ASTM Jet-A fuel. Full-scale JT9D experimental engine tests were conducted in a phase three aircraft experimental clean combustor program. The low-pollution combustor, fuel system, and fuel control concepts were derived from phase one and phase two programs in which several combustor concepts were evaluated, refined, and optimized in a component test rig. Significant pollution reductions were achieved with the combustor which meets the performance, operating, and installation requirements of the engine.

Roberts, R.↗
NASA NTRS1983

The NASA broad-specification fuels combustion technology program: An assessment of phase 1 test results

An assessment is made of the results of Phase 1 screening testing of current and advanced combustion system concepts using several broadened-properties fuels. The severity of each of several fuels-properties effects on combustor performance or liner life is discussed, as well as design techniques with the potential to offset these adverse effects. The selection of concepts to be pursued in Phase 2 refinement testing is described. This selection takes into account the relative costs and complexities of the concepts, the current outlook on pollutant emissions control, and practical operational problems.

Fear, J. S.↗
DOE OSTI2016

Key results from irradiation and post-irradiation examination of AGR-1 UCO Triso fuel - 18576

The AGR-1 irradiation experiment was performed as the first test of tri-structural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3 x 10{sup 5} coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time average, volume average irradiation temperatures of the individual compacts ranged from 955 to 1136 deg. C. This paper focuses on key results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior. The fuel exhibited zero TRISO coating failures (failure of all three dense coating layers) during irradiation and post-irradiation safety testing at temperatures up to 1700 deg. C. Advanced PIE methods have allowed particles with SiC coating failure that were discovered to be present in a very-low population to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of lessons learned from AGR-1 to fuel fabrication and post-irradiation examination for subsequent fuel irradiation experiments as part of the US fuel program are also discussed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Experimental Determination and Modeling of Used Fuel Drying by Vacuum and Gas Circulation for Dry Cask Storage

More than 120 drying tests were conducted using a mock fuel assembly with depleted uranium rods and heater rods to simulate decay heat. These tests followed standard industry practice for vacuum and forced helium drying (FHD). Both single effect tests evaluating a single fuel assembly or cask feature and combined effect tests were conducted. Single effect tests evaluated drying of a specific, known amount of water in a feature. Combined effect tests evaluated specific features and followed the flooding of the chamber, dewatering, and blowdown procedures before beginning the drying procedure. Specific simulated fuel assembly and canister features evaluated included PWR dashpot, BWR water rod, failed fuel rod, spacer disc, and Boral sheet. Ceria pellets were fabricated and used in the failed fuel rod to avoid issues of uranium contamination. A method of monitoring water content in the canister gas and gas stream was developed using optical emission spectroscopy (OES) in addition to monitoring of relative humidity in the chamber and in the vacuum lines.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2004

Benchmark Wall Heat Flux Data for a GO2/GH2 Single Element Injector

Computational Fluid Dynamics (CFD) is becoming an important component of injector design in the rocket industry. Injector designers who use CFD in the design process need to understand the accuracy level of the particular code being used for certain aspects of the design. This paper presents a recent effort to acquire benchmark quality data to be used for CFD code validation. Detailed chamber wall temperature and heat flux data was acquired for a gaseous oxygen, gaseous hydrogen single element shear coaxial injector in a 1.5 inch diameter copper heat sink chamber at Penn State University. The data was taken using both coaxial and water cooled heat flux gauges. Tests were run using hot gases generated from both fuel and oxidizer preburners. Tests were conducted over a chamber pressure range of 300 to 750 psia. Data analysis and uncertainty information will also be presented.

Pal, Sibtosh↗
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↗
DOE OSTI2025

Assessment of Buffer-IPyC Thermomechanical Debonding Behavior Using New Experiment Strength Data in BISON

TRIstructural ISOtropic (TRISO) single sub-millimeter-diameter nuclear fuel particle consists of a spherical fuel kernel surrounded by four coating layers. During reactor operation, the TRISO coating layers are subjected to irradiation-induced dimensional changes and the associated thermomechanical behavior of each layer. One of the observed behaviors is gap formation between the buffer and IPyC layer due to the porous buffer’s irradiation-induced shrinkage exceeding that of the IPyC layer. To predict TRISO thermomechanical performance, multi-physics models have been built that are continually updated and refined. The BISON code is a finite element-based nuclear fuel performance code that may be used for TRISO particle simulations. One of the recent additions to the BISON code is the ability to model the process of layer debonding. This paper will focus on the simulation results of the improved BISON debonding model that will utilize updated strengths measured from irradiated TRISO fuel particles. The new experimental strength data from micromechanical tests of irradiated TRISO fuel samples were exercised in the BISON simulations and compared to baseline strength data to assess their applicability in the models. This also includes updated buffer-IPyC bond strengths to simulate layer delamination. Based on current experimental observations it is noted that the buffer-IPyC separation occurs not exactly at the junction of these two layers, but more on the side of the buffer layer. This improved modeling approach using experimental strength data to characterize buffer-IPyC debonding and its potential subsequent cracking will be presented in the paper along with comparisons to available experimental observations.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Assessment of buffer-IPyC thermomechanical debonding behavior using new experimental strength data in BISON

TRIstructural ISOtropic (TRISO) fuel is a nuclear fuel commonly used in High Temperature Gas-cooled Reactors (HTGRs). A single sub-millimeter-diameter TRISO fuel particle consists of a spherical fuel kernel surrounded by four coating layers: a low-density pyrocarbon buffer layer, an inner pyrolytic carbon (IPyC) layer, a silicon carbide (SiC) layer, and an outer pyrolytic carbon (OPyC) layer. The kernel is commonly made of UO2 or a mixture of uranium carbide and uranium oxide (UCO). During reactor operation, the TRISO coating layers are subjected to irradiation-induced dimensional changes and the associated thermomechanical behavior of each layer. One of the observed behaviors is gap formation between the buffer and IPyC layer due to the porous buffer’s irradiation-induced shrinkage exceeding that of the IPyC layer. Not all irradiated particles will experience buffer-IPyC gap formation. The debonding may be partial, or it may be nearly total. However, from post-irradiation examination of UCO TRISO fuels irradiated as part of the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program, it was concluded that partial buffer-IPyC debonding was the most common type of buffer-IPyC interaction. To predict TRISO thermomechanical performance, multi-physics models have been built that are being continually updated and refined. The BISON code is a finite element-based nuclear fuel performance code that may be used for 1D, 2D, and 3D TRISO particle simulations. This code is used to calculate fuel temperature, kernel swelling, buffer densification, thermal and irradiation creep, fracture, and fission gas production and release. One of the recent additions to the BISON code is the ability to model the process of layer debonding. This paper will focus on the simulation results of the improved BISON debonding model that will utilize updated strengths measured from irradiated AGR TRISO fuel particles. The new experimental strength data from micromechanical tests of irradiated TRISO fuel samples were exercised in the BISON simulations and compared to baseline strength data to assess their applicability in the models. This also includes updated buffer-IPyC bond strengths to simulate layer delamination. Based on current experimental observations it is noted that the buffer-IPyC separation occurs not exactly at the junction of these two layers, but more on the side of the buffer layer. That observation is also implemented in the TRISO interface debonding model. This improved modeling approach using experimental strength data to characterize buffer-IPyC debonding and its potential subsequent cracking will be presented in the paper along with comparisons to available experimental observations.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2009

Nitric Oxide and Oxygen Air-Contamination Effects on Extinction Limits of Non-Premixed Hydrocarbon-Air Flames for a HIFiRE Scramjet

Unique nitric oxide (NO) and oxygen air-contamination effects on the extinction Flame Strength (FS) of non-premixed hydrocarbon (HC) vs. air flames are characterized for 7 gaseous HCs, using a new idealized 9.3 mm straight-tube Opposed Jet Burner (OJB) at 1 atm. FS represents a laminar strain-induced extinction limit based on cross-section-average air jet velocity, Uair, that sustains combustion of a counter jet of gaseous fuel just before extinction. Besides ethane, propane, butane, and propylene, the HCs include ethylene, methane, and a 64 mole-% ethylene / 36 % methane mixture, the writer s previously recommended gaseous surrogate fuel for HIFiRE scramjet tests. The HC vs. clean air part of the work is an extension of a May 2008 JANNAF paper that characterized surrogates for the HIFiRE project that should mimic the flameholding of reformed (thermally- or catalytically-cracked) endothermic JP-like fuels. The new FS data for 7 HCs vs. clean air are thus consolidated with the previously validated data, normalized to absolute (local) axial-input strain rates, and co-plotted on a dual kinetically dominated reactivity scale. Excellent agreement with the prior data is obtained for all 7 fuels. Detailed comparisons are also made with recently published (Univ. Va) numerical results for ethylene extinction. A 2009-revised ethylene kinetic model (Univ. Southern Cal) led to predicted limits within approx. 5 % (compared to 45 %, earlier) of this writer s 2008 (and present) ethylene FSs, and also with recent independent data (Univ. Va) obtained on a new OJB system. These +/- 5 % agreements, and a hoped-for "near-identically-performing" reduced kinetics model, would greatly enhance the capability for accurate numerical simulations of surrogate HC flameholding in scramjets. The measured air-contamination effects on normalized FS extinction limits are projected to assess ongoing Arc-Heater-induced "facility test effects" of NO production (e.g., 3 mole-%) and resultant oxygen depletion (from 21 to 19.5 %), for testing the "64/36" surrogate fuel in Langley s Arc-Heated Scramjet Test Facility for HIFiRE engine designs. The FS results show a generally small (< 4 %) "nitric oxide enhancement" effect, relative to clean air, for up to 3 % NO (freestream Mach number up to 7 in Arc Jet testing). However, a progressively large "oxygendeficiency weakening" effect develops. For 3 % NO, a net weakening of 26 % in FS is derived for the "64/36" fuel vs. air. The corresponding net weakening for pure ethylene is 20 %. A number of practical recommendations regarding facility test effects are offered.

Pellett, Gerald L.↗
DOE OSTI2016

Energy properties of solid fossil fuels and solid biofuels

The paper deals about the problematic of energy properties of solid biofuels in comparison with solid fossil fuels. Biofuels are alternative to fossil fuels and their properties are very similar. During the experiments were done in detail experiments to obtain various properties of spruce wood pellets and wheat straw pellets like biofuels in comparison with brown coal and black coal like fossil fuels. There were tested moisture content, volatile content, fixed carbon content, ash content, elementary analysis (C, H, N, S content) and ash fusion temperatures. The results show that biofuels have some advantages and also disadvantages in comparison with solid fossil fuels.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2017

AGR-5/6/7 Irradiation Test Predictions using PARFUME

PARFUME, (PARticle FUel ModEl) a fuel performance modeling code used for high temperature gas-cooled reactors (HTGRs), was used to model the Advanced Gas Reactor (AGR)-5/6/7 irradiation test using predicted physics and thermal hydraulics data. The AGR-5/6/7 test consists of the combined fifth, sixth, and seventh planned irradiations of the AGR Fuel Development and Qualification Program. The AGR-5/6/7 test train is a multi-capsule, instrumented experiment that is designed for irradiation in the 133.4-mm diameter north east flux trap (NEFT) position of Advanced Test Reactor (ATR). Each capsule contains compacts filled with uranium oxycarbide (UCO) unaltered fuel particles. This report documents the calculations performed to predict the failure probability of tristructural isotropic (TRISO)-coated fuel particles during the AGR-5/6/7 experiment. In addition, this report documents the calculated source term from the driver fuel. The calculations include modeling of the AGR-5/6/7 irradiation that is scheduled to occur from October 2017 to April 2021 over a total of 13 ATR cycles, including nine normal cycles and four Power Axial Locator Mechanism (PALM) cycle for a total between 500 – 550 effective full power days (EFPD). The irradiation conditions and material properties of the AGR-5/6/7 test predicted zero fuel particle failures in Capsules 1, 2, and 4. Fuel particle failures were predicted in Capsule 3 due to internal particle pressure. These failures were predicted in the highest temperature compacts. Capsule 5 fuel particle failures were due to inner pyrolytic carbon (IPyC) cracking causing localized stresses concentrations in the SiC layer. This capsule predicted the highest particle failures due to the lower irradiation temperature. In addition, shrinkage of the buffer and IPyC layer during irradiation resulted in formation of a buffer-IPyC gap. The two capsules at the two ends of the test train, Capsules 1 and 5 experienced the smallest buffer-IPyC gap formation due to the lower irradiation fluences and temperatures. Capsule 3 experienced the largest buffer-IPyC gap formation of just under 24 µm. The release fraction of fission products Ag, Cs, and Sr silver (Ag), cesium (Cs), and strontium (Sr) vary depending on capsule location and irradiation temperature. The maximum release fraction of Ag occurs in Capsule 3, reaching up to 84.8% for the TRISO fuel particles. The release fraction of the other two fission products, Cs and Sr are much smaller and, in most cases, less than 1%. The notable exception is again in Capsule 3, where the release fraction for Cs and Sr reach up to 9.7% and 19.1%, respectively.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Autonomous Fueling System for Heavy-Duty Fuel Cell Electric Trucks

The motivation for this project stemmed from the challenges associated with rapidly refueling heavy-duty hydrogen fuel cell electric trucks (FCETs). Current manual refueling processes for fast refueling involve large, heavy equipment (e.g., hoses three times heavier than standard) and pose ergonomic risks and potential for equipment damage. The goal was to develop and test an autonomous fueling system to improve ergonomics, enhance safety, increase equipment durability through design improvements, and potentially speed up the fueling process. This project aimed to add to the understanding of autonomous systems in the context of heavy-duty hydrogen refueling, evaluating the technical effectiveness of potential concepts. A successfully developed system would benefit the public by facilitating the adoption of zero-emission heavy-duty transport, reducing reliance on manual labor for a physically demanding task, and potentially improving the safety and efficiency of hydrogen refueling infrastructure. The major accomplishment during the project's active period was the completion of the system-level architecture task. This involved establishing a detailed list of system requirements covering interfaces, environmental conditions, regulatory compliance, industry standards, safety, security, performance capabilities, and optional features. Five key use cases for the autonomous system were also identified. However, due to internal restructuring at Nikola, the necessary resources could not be allocated to continue the project. Consequently, Nikola opted to discontinue the project. The award was mutually terminated by Nikola and the DOE.

08 HYDROGEN↗
NASA NTRS2013

Lean Development with the Morpheus Simulation Software

The Morpheus project is an autonomous robotic testbed currently in development at NASA's Johnson Space Center (JSC) with support from other centers. Its primary objectives are to test new 'green' fuel propulsion systems and to demonstrate the capability of the Autonomous Lander Hazard Avoidance Technology (ALHAT) sensor, provided by the Jet Propulsion Laboratory (JPL) on a lunar landing trajectory. If successful, these technologies and lessons learned from the Morpheus testing cycle may be incorporated into a landing descent vehicle used on the moon, an asteroid, or Mars. In an effort to reduce development costs and cycle time, the project employs lean development engineering practices in its development of flight and simulation software. The Morpheus simulation makes use of existing software packages where possible to reduce the development time. The development and testing of flight software occurs primarily through the frequent test operation of the vehicle and incrementally increasing the scope of the test. With rapid development cycles, risk of loss of the vehicle and loss of the mission are possible, but efficient progress in development would not be possible without that risk.

Brogley, Aaron C.↗
NASA NTRS2001

NASA Multipoint LDI Development

Multipoint Lean-Direct-Injection (LDI) is a combustor concept in which a large number of fuel injectors and fuel-air mixers are used to quickly and uniformly mix the fuel and air so that ultralow levels of NO, are produced. Each fuel injector has an air swirler associated with it for fuel-air mixing and to establish a small recirculation and burning zone. A concept in which there are 36 fuel injectors in the space of a conventional single fuel injector has been tested in a flame tube. A greater than 80 percent reduction in NO, at high power conditions (400 psia, 1000 "Finlet) was achieved. Alternate concepts with 9,25,36 or 49 fuel injectors are being investigated in flame tube tests for their low NO, potential and with fuel staging to improve the turn-down ratio at low power conditions. A preliminary sector concept of a large engine design has been successfully tested at inlet conditions of 700 psia and 1100 O F . This concept had one half the number of fuel injectors per square inch as the flame tube configuration with 36 fuel injectors, and the NO, reduction was 65 percent of the ICAO standard. Future regional engine size sector tests are planned for the 2nd quarter of FY02 and large engine size sector tests for the 1st quarter of FY03.

Tacina, Robert↗