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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 595 records · Page 33
All sourcesNASA NTRSDOE OSTI
NASA NTRS2011

Characterization of Deficiencies in the Frequency Domain Forced Response Analysis Technique for Supersonic Turbine Bladed Disks

Turbine blades in rocket and jet engine turbomachinery experience enormous harmonic loading conditions. These loads result from the integer number of upstream and downstream stator vanes as well as the other turbine stages. Assessing the blade structural integrity is a complex task requiring an initial characterization of whether resonance is possible and then performing a forced response analysis if that condition is met. The standard technique for forced response analysis in rocket engines is to decompose a CFD-generated flow field into its harmonic components, and to then perform a frequency response analysis at the problematic natural frequencies. Recent CFD analysis and water-flow testing at NASA/MSFC, though, indicates that this technique may miss substantial harmonic and non-harmonic excitation sources that become present in complex flows. A substantial effort has been made to account for this denser spatial Fourier content in frequency response analysis (described in another paper by the author), but the question still remains whether the frequency response analysis itself is capable of capturing the excitation content sufficiently. Two studies comparing frequency response analysis with transient response analysis, therefore, of bladed-disks undergoing this complex flow environment have been performed. The first is of a bladed disk with each blade modeled by simple beam elements. Six loading cases were generated by varying a baseline harmonic excitation in different ways based upon cold-flow testing from Heritage Fuel Air Turbine Test. It was hypothesized that the randomness and other variation from the standard harmonic excitation would reduce the blade structural response, but the results showed little reduction. The second study was of a realistic model of a bladed-disk excited by the same CFD used in the J2X engine program. It was hypothesized that enforcing periodicity in the CFD (inherent in the frequency response technique) would overestimate the response. The results instead showed that the transient analysis results were up to 10% higher for "clean" nodal diameter excitations and six times larger for "messy" excitations, where substantial Fourier content around the main harmonic exists. Because the bulk of resonance problems are due to the "clean" excitations, a 10% underprediction is not necessarily a problem, especially since the average response in the transient is similar to the frequency response result, and so in a realistic finite life calculation, the life would be same. However, in the rare cases when the "messy" excitations harmonics are identified as the source of potential resonance concerns, this research does indicate that frequency response analysis is inadequate for accurate characterization of blade structural capability.

Brown, Andrew M.↗
DOE OSTI2021

Fuel Design and Fabrication: Research Reactor Fuel

The objective of this chapter is to inform the reader on the essential aspects of design and fabrication of nuclear fuel systems utilized in research and test reactors. Fuel attributes of interest include mechanical, physical, and thermal properties as well as overall manufacturability. The chapter focuses primarily on plate-type fuels utilized in the highest power research reactors where design and function are critical to operational safety and successfully meeting each reactor’s unique mission.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Scaling for Nuclear Reactor System: Overview

A scaling analysis is essential for predicting the behavior of a reactor system under various operating conditions, as well as for ensuring safe and reliable experimentation and adequate dataset generation for supporting the system code model development and assessment in accordance with USNRC guidelines 1.203. Testing and modeling can be carried out at various scales, from small test units to full-scale reactors, to validate the predictions produced by the analysis. However, due to the cost and challenges associated with full-scale testing, the nuclear industry uses scaled test facilities—integral effect test (IET) and separate effect test (SET)—to identify potential design issues and optimize the design to ensure regulatory compliance for safer operation and anticipated accident management. In general, IET facilities are non-nuclear thermal hydraulics test facilities. In contrast, SET facilities can be non-nuclear testing (e.g., steam generator, containment, pump system SETs) and nuclear testing (mostly fuel-material irradiation testing). IETs are used to study the system-level thermal-hydraulic behavior of the reactor coolant system (RCS) under different anticipated operational occurrences (AOOs) and design basis accidents (DBAs) such as loss-of-coolant-accident (LOCA) and loss-of-power (i.e., station blackout) and loss-of-flow (i.e., failure of RCS pump). Several LOCA scenarios could happen in reactor DBAs, such as main steam line break (MSLB), SG tube rapture (SGTR), break in direct vessel injection (DVI), and other small breaks LOCA. Scaling analysis determines the appropriate size and operating conditions for the test facility. This involves analyzing the system's thermal-hydraulic behavior at different scales and identifying the key parameters that affect the thermal-hydraulic behavior of the reactor. These key parameters can be identified so that the scaled facility can replicate the expected conditions in a prototypical setting. The scaled facility can simulate the same flow rates, pressure gradients, heat transfer coefficients, and fluid properties as the prototype. Scaling analysis is used to ensure that the results from these tests can be scaled up to predict the behavior of a full-scale prototype with an acceptable range of scaling distortions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2004

RQL Fuel Shifting Sector Rig Test

The low emissions potential of a Rich-Quench-Lean (RQL) combustor for use in the HIgh Speed Civil transport (HSCT) application was evaluated as part of the NASA Critical Propulsion Components (CPC) Program. Fuel shifting as an approach to combustor control was evaluated in a multiple bank RQL combustor, utilizing reduced scale quench technology implemented in a convoluted linear with quench plate concept.

Holdeman, James D.↗
DOE OSTI2017

Development of High Temperature Out-of-Pile Experiments for Testing Nuclear Thermal Propulsion Fuel Surrogates

Nuclear thermal propulsion (NTP) technology is designed to provide “in-space” vehicle propulsion by deriving thrust from a hydrogen working fluid that is heated from a nuclear reaction. Initial NTP fuel structures (primarily carbon/carbide based materials) developed by the Rover/NERVA program were designed to withstand temperatures exceeding 2400 K and remain intact after tens of firing cycles. Current programs responsible for development of this technology expect the fuels to withstand temperatures in excess of 2550K under a similar duty cycle. Experiments that subject fuels to these conditions are being developed to provide data and performance information for optimization and potential down-selection of the candidate materials. Initial tests will be designed for laboratory benchtop operation and will not expose surrogate specimens to the hydrogen working fluid or irradiation fields. The primary focus of these out-of-pile experiments will be to study thermal performance and thermal failure limits of the candidate specimen materials.

Howard, Richard↗
DOE OSTI2021

Preliminary results from Low Pressure Steam Oxidation Testing of ALD ZrN and ZrO2 Coating Deposited over UCN Fuel Kernels

Steam oxidation testing was used to investigate as-developed 100 nm ALD ZrN and 1000 nm ZrO 2 coatings deposited over UC1-xNx fuel kernels. The results of these tests were compared against oxidation of uncoated kernels. This work was performed to support development of coatings which both provide high temperature Zr metal diffusion barrier and also provide resistance against oxidation, especially against high temperature steam and/or air. This work was performed in collaboration with Oak Ridge National Laboratory (ORNL), using ORNL-provided samples. The oxidation tests yielded valuable information, including relative differences in performance of different coating materials and coating thicknesses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Thermal-hydraulic and Fuel Performance Scoping Studies of a Flowing Water Capsule in TREAT

The restart of the Transient Reactor Test (TREAT) facility has provided a much needed capability for integral transient testing of nuclear fuel. This testing is necessary to qualify new fuel concepts such as those developed under the Accident Tolerant Fuel (ATF) program, increasing the burnup limit of light water reactor (LWR) fuels, or a variety of other programs under the Advanced Fuels Campaign (AFC). The ATF campaign has been the main driver behind the development and implementation of a variety of capsules for TREAT experiments. The Separate-Effect Test Holder (SETH) was a dry capsule that enabled testing of ATF concepts under reactivity-initiated accident (RIA) heating conditions. Following SETH, the Static Environment Rodlet Transient Test Apparatus (SERTTA) was developed to enable RIA testing in a static water environment. In efforts to support the need for future Loss-of-Coolant Accident (LOCA) tests, the Transient Water Irradiation System for TREAT (TWIST) capsule has been developed that allows for water to drain from around the fuel rod and lower the pressure to simulate LOCA conditions. All these capsules that have been developed for LWR fuel testing all lack the capability for forced convective cooling which in some applications may limit their ability to test under prototypic conditions. A design and modeling effort has been started to modify the TWIST capsule by adding a flow tube and impeller that can create flowing coolant conditions for the fuel rod. RELAP5-3D and BISON models are being used to study the differences between RIA, LOCA, Anticipated Operational Occurrences (AOO), LWR power cycling, and other transient scenarios under flowing conditions capable in the flowing capsule and the current stagnant water capsules (SERTTA and TWIST). The scoping study will provide guidance on the needed capabilities for the flowing capsule and the limitations of the currently developed capsule for LWR testing in TREAT.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Status of Triso fuel irradiations in the advanced test reactor supporting high-temperature gas-cooled reactor designs - 18468

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven experiments on low-enriched uranium tri-structural isotropic (TRISO) particle fuel. Irradiations and fuel development are being accomplished at Idaho National Laboratory s Advanced Test Reactor to support development of the next generation of reactors in the United States. The experiments are being irradiated over several years to demonstrate and qualify new TRISO-coated particle fuel for use in high-temperature gas reactors, with goals of providing irradiation performance data to support fuel process development, qualifying fuel for normal operating conditions, supporting development and validation of fuel performance and fission product transport models and codes, and providing irradiated fuel and materials for post-irradiation examination and safety testing. The experiments are irradiated in an inert sweep gas atmosphere, with individual online temperature monitoring and control of each capsule and online fission product monitoring of effluent to track individual capsule fuel performance. The first experiment (AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Because the purpose of this experiment was to provide data on fission product migration and retention in the next generation reactor, the design of this experiment was significantly different from the first two experiments. The final experiment (AGR-5/6/7) is scheduled to begin irradiation in early summer 2017. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

High-Activity Dealloyed Catalysts

Reduction of costly Pt usage in proton exchange membrane fuel cell electrodes is one of the major challenges towards development and commercialization of fuel cell vehicles. Although few have met the initial-kinetic activity requirements in a realistic fuel cell device, no catalyst material has ever met the demanding fuel cell durability targets set by DOE. In this project, a team of 4 universities and 2 companies came together to investigate a concept that appeared promising in preliminary non-fuel cell tests then to further develop the catalyst to a mature level ready for vehicle implementation. The team consists of academia with technical leadership in their respective areas, a catalyst supplier, and a fuel cell system integrator.The tightly collaborative project enabled development of a highly active and durable catalyst with performance that significantly exceeds that of previous catalysts and meets the DOE targets for the first time (Figure 1A). The catalyst was then further evaluated in full-active-area stack in a realistic vehicle operating condition (Figure 1B). This is the first public demonstration that one can realize the performance benefit and Pt cost reduction over a conventional pure Pt catalyst in a long-term realistic PEMFC system. Furthermore, systematic analyses of a range of catalysts with different performance after fuel cell testing allowed for correlation between catalyst microstructure and its electrocatalytic activity and durability. This will in turn aid future catalyst development.

30 DIRECT ENERGY CONVERSION↗
NASA NTRS1981

Autoignition characteristics of aircraft-type fuels

An investigation was conducted of the ignition delay characteristics of five liquid hydrocarbon fuels in air. The test apparatus developed permitted independent variation and control of temperature, pressure, air flow rate, and fuel/air ratio in order that the effects of each parameter could be investigated independently. All of the fuels tested behaved in a predictable manner, that is, ignition delay time decreased as temperature, pressure, and fuel/air ratio increased. The results for the different fuels tested (Jet-A, MP-4, No. 2 diesel, ERBS, and cetane) were directly comparable, since it can be shown that the fuel spray characteristics from the multiple conical tube injector are relatively insensitive to small changes in fuel properties (viscosity, surface tension, and density).

Spadaccini, L. J.↗
DOE OSTI2017

Multivariate analysis of gamma spectra to characterize used nuclear fuel

Abstract—The Multi-Isotope Process (MIP) Monitor provides an efficient approach to monitoring the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of reprocessing streams in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor), initial enrichment, burn up, and cooling time. Simulated gamma spectra were used to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type. Locally weighted PLS models were fitted on-the-fly to estimate continuous fuel characteristics. Burn up was predicted within 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment within approximately 2% RMSPE. This automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters and material diversions.

Multivariate Analysis for Characterization of Used↗
NASA NTRS1979

Compatibility of elastomers in alternate jet fuels

The compatibility of elastomeric compositions of known resistance to aircraft fuels was tested for potential use in Jet A type fuels obtainable from alternate sources, such as coal. Since such fuels were not available at the time, synthetic alternate fuels were prepared by adding tetralin to a petroleum based Jet A type fuel to simulate coal derived fuels which are expected to contain higher amounts of aromatic and hydroaromatic hydrocarbons. The elastomeric compounds tested were based on butadiene-acrylonitrile rubber, a castable Thiokol polysulfide rubber, and a castable fluorosilicone rubber. Batches of various cross-link densities of these rubbers were made and their chemical stress relaxation behavior in fuel, air, and nitrogen, their swelling properties, and response to mechanical testing were determined.

Kalfayan, S. H.↗
NASA NTRS2003

Putting Fuel Cells to the Test

If research has its way, an electrochemical device capable of converting energy into electricity and heat will become the impetus behind the next generation of automobiles, superseding the internal combustible engine found under the hoods of vehicles that rule the road today. The thought of fuel cell technology being able to accomplish such a feat may be dismissed as too futuristic by some, but the truth is that fuel cells have been in play as a source of propulsion since the 1960s, when NASA first used them to generate power onboard the Gemini and Apollo spacecraft for extended space missions. Even more unknown is the fact that fuel cells were and continue to be a source of drinking water for astronauts in orbit, since they produce pure water as a by-product.

Source record↗
DOE OSTI2023

Accelerated test protocols to predict service life and durability of solid oxide fuel cells

Reliable accelerated test protocols are needed for solid oxide fuel cell research to facilitate rapid learning on key durability issues, identify potential modes of failure expeditiously, and eventually predict the calendar lifetime of an electrochemical cell. In this work, solid oxide fuel cells operated at a constant current density were compared to cells undergoing accelerated measurements, which are composed of intermittent current injection to the cell. A general accelerated test profile was developed by cycling a solid oxide fuel cell from open circuit to a predetermined operating current density that is the same as the current density during a steady-state operation, to accelerate the local redox environment. The following parameters were studied: current density, operation temperature, moist level, sintering temperature, cycling current, cycling frequency, and operation time. Up to 1,320,000 cycles were generated in this work. The cell degradation was accelerated by nearly 10 times, suggesting the feasibility of using this protocol for acceleration test to predict life performance and durability of solid oxide fuel cells.

08 HYDROGEN↗
DOE OSTI2023

Design and testing of an enriched uranium fueled molten salt irradiation vehicle

Molten salt reactors (MSRs) have garnered increasing attention recently with several demonstration efforts on the way. A key challenge to the licensing basis for these reactors is the lack of experimental data on fueled salts. This is expected to be crucial to the safety evaluation and licensing basis of reactors of this type deployed in the future. While capability for irradiating molten salts has started being reestablished in the recent decade, no enriched fuel irradiation capability has been developed and tested as of yet. A new experiment vehicle under development at Idaho National Laboratory (INL) is presented here. The Molten-salt Research Temperature-controlled Irradiation (MRTI) experiment was developed to host enriched-uranium bearing salt samples to be irradiated at a test reactor within the lab complex. One of the key scientific objectives is to provide irradiated salt samples for post irradiation examination (PIE) to study the impact of fission product generation and neutron/gamma radioactivity on the salt solution and salt-facing wall material. This paper provides a detailed overview of the mechanical design of the experiment, followed by an overview of the fabrication and assembly of an initial prototype vehicle (with non-fuel bearing salt). A summary of the key analyses conducted as a part of the performance and safety evaluation is then provided. Lastly, an overview of the test conducted in prototypic out-of-pile (non-neutron) environment are shown. These evaluations provide the foundation for a planned irradiation of and enriched uranium-bearing chloride salt sample in the near term. The upcoming irradiation will contain 13cm 3 of UCl 3 -NaCl salt (93% enrichment) generating around 20 W/cm 3 of fission energy during irradiation and a temperature range that can be contained between bounds of 525-900°C.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Radial Deconsolidation Leach-Burn-Leach of Five AGR-3/4 TRISO Fuel Compacts after Post-Irradiation Heating Tests

Five AGR-3/4 fuel compacts were previously heated in the Fuel Accident Condition Simulator (FACS) furnace. Three of these compacts were also reirradiated in the Neutron Radiography (NRAD) reactor before the FACS tests. One of these was heated to 1200°C, two were heated to 1400°C, one was heated to 1600°C, and one was held for a period of time at 1600 and 1700°C. After these FACS tests, the compacts were subjected to destructive post-irradiation examination (PIE) via radial-deconsolidation-leach-burn-leach (RDLBL) at INL. The RDLBL process deconsolidated the compacts in multiple radial steps, followed by a final, single-step axial deconsolidation. The samples generated at each step were analyzed for isotopes of key fission products and actinides. After each deconsolidation step, the compact volume was assessed, and this was used to normalize the measured quantities of nuclides of interest to give volumetric concentrations as a function of radial position within the compact. The total inventories of measured fission products and actinides and the radial concentration profiles were compared with their sibling compacts within the same capsule (similar irradiation conditions) that went through RDLBL in the as-irradiated state. These data will be used to refine the fission product transport models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Post-irradiation Heating Tests of As-Irradiated AGR-3/4 TRISO Fuel Compacts

Four post-irradiation heating tests of fuel compacts from the U.S. Advanced Gas Reactor (AGR)-3/4 irradiation experiment were completed. In addition to tristructural isotropic (TRISO)-coated driver fuel, each compact contained designed-to-fail (DTF) particles with fuel kernels coated only in pyrocarbon so as to simulate exposed kernels. Tests at 1600/1700°C, 1400°C, and 1200°C were performed to measure fission product release as a function of time and temperature. Silver release was highest in the 1200°C test, supporting the observation that silver release rates are highest in the 1100–1300°C range. Compared to tests of AGR-1 compacts with no exposed kernels, the Cs-134 and Kr-85 releases were noticeably higher in AGR-3/4. The exposed kernels’ contributions to Eu and Sr release are inconclusive, due to the difficulty in distinguishing among the combined effects of higher irradiation temperatures in these particular AGR-3/4 compacts, the presence of the DTF particles, and the Fuel Accident Condition Simulator (FACS) test temperatures. These data can be used to make inferences about fission product retention in exposed kernels as a function of time and temperature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Reducing Uncertainty in Hydrodynamic Modeling of ATR Experiments Via Flow Testing, Validation, and Optimization

The development, characterization, and qualification testing of nuclear fuel at Idaho National Laboratory’s Advanced Test Reactor (ATR) requires extensive design and analysis activities prior to the insertion of an irradiation experiment in-pile. Significant effort is made in the design and development phase of all in-pile experiments to ensure that the maximum feasible impacts of all necessary experimental requirements are satisfied. The advancement of fuel, cladding, and in-reactor materials technology in recent years has introduced complexities associated with the design and construct of in-pile experiments necessitating deeper understanding of boundary conditions and increasingly comprehensive observations resulting from the experiment. Each unique experiment must be assessed for neutronics response, thermal/hydraulic/hydrodynamic performance, and structural integrity. This is accomplished either analytically, computationally, or experimentally, or some combination thereof, prior to insertion into the ATR. The various effects are interrelated to various degrees, such as the case with the experiment temperature affecting the thermal cross section of the fuel or the increased temperature of the experiment’s materials reducing the mechanical strength of the assemblies. Additionally, the feedback between the experiment’s response to a reactor transient could alter the neutron flux profile of the reactor during the transient. Each experiment must therefore undergo a barrage of analyses to assure the ATR operational safety review committee that the insertion and irradiation of the experiment will not detrimentally affect the safe operational envelope of the reactor. In many cases, the nuclear fuel being tested can be double-encapsulated to ensure safety margins are adequately addressed, whereas failed fuel would be encased in a protective capsule. In other cases, the experiments can be inserted in a self-contained loop that passes through the reactor core, remaining isolated from the primary coolant. In the case of research reactor fuel, however, the fuel plates must be tested in direct contact with the reactor coolant, and being fuel designed for high neutron fluxes, they are inherently power-dense plates. The combination of plate geometry, high-power density, and direct contact with primary coolant creates a scenario where the neutronic/thermomechanic/hydrodynamic characteristics of the fuel plates are tightly coupled, necessitating as complete characterization as possible to support the safety and programmatic assessments, thus enabling a successful experiment. This paper explores the efforts of the U.S. High-Performance Research Reactor program to thermomechanically/hydromechanically characterize the program’s wide variety of experiments, which range from stacks of miniplate capsules to full-sized, geometrically representative curved plates. Special attention is given to instances where the combination of experimental characterization and analytical assessment has reduced uncertainties of the safety margins, allowing experiments to be irradiated that would otherwise not have passed the rigorous qualification process for irradiation in the ATR. In some cases, the combined processes have exposed flow and heat transfer characteristics that would have been missed using historical methods, which allows for more accurate and representative postirradiation assessments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗