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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 1,045 records · Page 58
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Understanding TRISO Coated Particle Neutron Irradiation Behavior:Evolution of Advanced Micro Analysis and Electron Microscopy Approaches

Objectives and motivation• Provide data for fuel qualification in support of reactor licensing• Establish a domestic commercial TRISO fuel fabrication capability Approach • Focus is on developing and testing UCO TRISO fuel –Develop fuel fabrication and QC measurement methods, first at lab scale and then at industrial scale –Perform irradiation testingover a range of conditions (burnup, temperature, fast neutron fluence) –Perform post-irradiation examination and safety testingto demonstrate and understand performance during irradiation and during accident conditions –Develop fuel performance modelsto better predict fuel behavior –Perform fission product transport experiments to improve understanding and refine models of fission product transport

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1971

Out-of-core evaluations of a nonfueled and a UO2-fueled cylindrical thermionic converter.

Two cylindrical thermionic energy converters similar to those in a flashlight thermionic fuel element were tested electrically at the Jet Propulsion Laboratory. One converter was not fueled, but the other was fueled with UO2 imbedded in six pencil-lead-size holes in the emitter, which was made of rhenium. The nonfueled converters showed no change in its performance during 4000 hr of testing, whereas the output current of the UO2-fueled converter degraded 15% during 2400 hr of testing at 2000 K. Measurements on the UO2-fueled converter showed an increase in the collector work function and an increase in the bare emitter work function. The degradation appears to be caused primarily by foreign deposits on the collector, probably uranium that diffused from the UO2 fuel through the emitter.

Shimada, K.↗
NASA NTRS2008

Analysis of the Effects of Vitiates on Surface Heat Flux in Ground Tests of Hypersonic Vehicles

To achieve the high enthalpy conditions associated with hypersonic flight, many ground test facilities burn fuel in the air upstream of the test chamber. Unfortunately, the products of combustion contaminate the test gas and alter gas properties and the heat fluxes associated with aerodynamic heating. The difference in the heating rates between clean air and a vitiated test medium needs to be understood so that the thermal management system for hypersonic vehicles can be properly designed. This is particularly important for advanced hypersonic vehicle concepts powered by air-breathing propulsion systems that couple cooling requirements, fuel flow rates, and combustor performance by flowing fuel through sub-surface cooling passages to cool engine components and preheat the fuel prior to combustion. An analytical investigation was performed comparing clean air to a gas vitiated with methane/oxygen combustion products to determine if variations in gas properties contributed to changes in predicted heat flux. This investigation started with simple relationships, evolved into writing an engineering-level code, and ended with running a series of CFD cases. It was noted that it is not possible to simultaneously match all of the gas properties between clean and vitiated test gases. A study was then conducted selecting various combinations of freestream properties for a vitiated test gas that matched clean air values to determine which combination of parameters affected the computed heat transfer the least. The best combination of properties to match was the free-stream total sensible enthalpy, dynamic pressure, and either the velocity or Mach number. This combination yielded only a 2% difference in heating. Other combinations showed departures of up to 10% in the heat flux estimate.

Cuda, Vincent↗
DOE OSTI2014

Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the High Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1981

Development of a high-temperature durable catalyst for use in catalytic combustors for advanced automotive gas turbine engines

Durable catalytic reactors for advanced gas turbine engines were developed. Objectives were: to evaluate furnace aging as a cost effective catalytic reactor screening test, measure reactor degradation as a function of furnace aging, demonstrate 1,000 hours of combustion durability, and define a catalytic reactor system with a high probability of successful integration into an automotive gas turbine engine. Fourteen different catalytic reactor concepts were evaluated, leading to the selection of one for a durability combustion test with diesel fuel for combustion conditions. Eight additional catalytic reactors were evaluated and one of these was successfully combustion tested on propane fuel. This durability reactor used graded cell honeycombs and a combination of noble metal and metal oxide catalysts. The reactor was catalytically active and structurally sound at the end of the durability test.

Tong, H.↗
NASA NTRS1985

Onsite 40-kilowatt fuel cell power plant manufacturing and field test program

A joint Gas Research Institute and U.S. Department of Energy Program was initiated in 1982 to evaluate the use of fuel cell power systems for on-site energy service. Forty-six 40 kW fuel cell power plants were manufactured at the United Technologies Corporation facility in South Windsor, Connecticut, and are being delivered to host utilities and other program participants in the United States and Japan for field testing. The construction of the 46 fully-integrated power plants was completed in January 1985 within the constraints of the contract plan. The program has provided significant experience in the manufacture, acceptance testing, deployment, and support of on-site fuel cell systems. Initial field test results also show that these experimental power plants meet the performance and environmental requirements of a commercial specification. This Interim Report encompasses the design and manufacturing phases of the 40 kW Power Plant Manufacturing and Field Test program. The contract between UTC and NASA also provides UTC field engineering support to the host utilities, training programs and associated manuals for utility operating and maintenance personnel, spare parts support for a defined test period, and testing at UTC of a power plant made available from a preceding program phase. These activities are ongoing and will be reported subsequently.

Source record↗
NASA NTRS2002

Compatibility Studies of Hydrogen Peroxide and a New Hypergolic Fuel Blend

Several preliminary materials compatibility studies have been conducted to determine the practicality of a new hypergolic fuel system. Hypergolic fuel ignites spontaneously as the oxidizer decomposes and releases energy in the presence of the fuel. The bipropellant system tested consists of high-test hydrogen peroxide (HTP) and a liquid fuel blend consisting of a hydrocarbon fuel, an ignition enhancer and a transition metal catalyst. In order for further testing of the new fuel blend to take place, some basic materials compatibility and HTP decomposition studies must be accomplished. The thermal decomposition rate of HTP was tested using gas evolution and isothermal microcalorimetry (IMC). Materials were analyzed for compatibility with hydrogen peroxide including a study of the affect welding has on stainless steel elemental composition and its relation to HTP decomposition. Compatibility studies of valve materials in the fuel blend were performed to determine the corrosion resistance of the materials.

Baldridge, Jennifer↗
DOE OSTI2022

Application of Dynamical System Scaling for Accelerated Fuel Qualification Efforts

There are considerable ongoing research and testing campaigns to qualify new fuel designs such as metallic fuels for advanced reactor designs and the Accident Tolerant Fuel (ATF) campaign for light water reactors (LWR). The typical research and development lifecycle needed to qualify a nuclear fuel design, under ideal conditions, can take up to 20-25 years which limits the ability for new fuels to make fast deployments into commercial, test, and research reactors. While there are several efforts to accelerate nuclear fuel qualification through advanced modeling coupled with state-of-the-art experiments, there is not yet a framework to methodically quantify and rank distortions occurring in experimental test specimens needed to validate nuclear fuel performance codes. This study uses the Dynamical System Scaling (DSS) methodology to quantify transient scaling distortions that occur over experiment and simulated transients. DSS analysis metrics provide a researcher with several tools and information embedded within test data to identify dominant phenomena, associated timescales, and quantify a transient?s overall dynamic distortion. This is demonstrated with separate analyses of the SETH-C and CHF-SERTTA thermal-hydraulic experiments in the TREAT facility at Idaho National Laboratory. The outcome of this work is a scaling and data analysis approach to account for the influence a scaled fuel test specimen?s geometric and temporal distortions have on its ability to be representative of the full-scale design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Evolution of pores in the fuel cladding of the electrogenerating channel

The results of reactor tests of carbonitride fuel in a monocrystalline cladding from a molybdenum-based alloy that were carried out earlier on an experimental setup Ya-82 for 8 to 300 h at a temperature of ∼1500°C can be used in order to prove the operational reliability of fuel elements in the design of a megawatt nuclear power plant for a spacecraft. A raster image of the surface of a sample of cladding shows that the interfaces between the layers are decorated by pores. This result is explained in this work by the theory of coalescence. The mechanisms responsible for the evolution of pores taking place at the parameters of a Ya-82 setup are considered. The effect of decoration by pores of the interfaces between the layers of a sample of cladding of the electrogenerating channel by carrying out reactor tests is explained. The dependence of the average radius of pores on the duration of the experiment is obtained. An evaluation of the average sizes of pores arising under the conditions of the experiment gives a value of ∼2 μm, which is in agreement with the experimental data. A computational study of swelling of the cladding material in the process of irradiation is performed. Predictive estimates of the behavior of the porous system and swelling of the cladding material for a megawatt class nuclear power plant are made.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Bending testing and characterization of surrogate nuclear fuel rods made of Zircaloy-4 cladding and aluminum oxide pellets

We studied behavior of surrogate nuclear fuel rods made of Zircaloy-4 (Zry-4) cladding with alumina pellets under reversed cyclic bending. Tests were performed under load or moment control at 5 Hz, and an empirical correlation was established between rod fatigue life and amplitude of the applied moment. Fatigue response of Zry-4 cladding was further characterized by using flexural rigidity. Degradation of flexural rigidity was shown to depend on the moment applied and the prefatigue condition of specimens. Pellet-to-pellet interface (PPI), pellet-to-cladding interface (PCI), and pellet condition all affect surrogate rod failure. Bonding/debonding of PPI/PCI and pellet fracturing contribute to surrogate rod bending fatigue. Also, the effect of sensor spacing on curvature measurement using three-point deflections was studied; the method based on effective specimen gauge length is effective in sensor spacing correction. Finally, we developed the database and gained understanding in this study such that it will serve as input to analysis of SNF vibration integrity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

New Fiber Materials with Sorption Capacity at 5.0 g-U/kg Adsorbent under Marine Testing Conditions

The Fuel Resources program of the Fuel Cycle Research and Development program of the Office of Nuclear Energy (NE) has focused on assuring that nuclear fuel resources are available in the United States for a long term. An immense source of uranium is seawater, which contains an estimated amount of 4.5 billion tonnes of dissolved uranium. Extraction of the uranium resource in seawater can provide a price cap and ensure centuries of uranium supply for future nuclear energy production. NE initiated a multidisciplinary program with participants from national laboratories, universities, and research institutes to enable technical breakthroughs related to uranium recovery from seawater. The goal is to develop advanced adsorbents to make the seawater uranium recovery technology a cost competitive, viable technology. Under this program, Oak Ridge National Laboratory (ORNL) has developed several novel adsorbents, which enhanced the uranium capacity 4-5 times from the state-of-the art Japanese adsorbents. Uranium exists uniformly at a concentration of ~3.3 ppb in seawater. Because of the vast volume of the oceans, the total estimated amount of uranium in seawater is approximately 1000 times larger than its amount in terrestrial resources. However, due to the low concentration, a significant challenge remains for making the extraction of uranium from seawater a commercially viable alternative technology. The biggest challenge for this technology to overcome to efficiently reduce the extraction cost is to develop adsorbents with increased uranium adsorption capacity. Two major approaches were investigated for synthesizing novel adsorbents with enhanced uranium adsorption capacity. One method utilized conventional radiation induced graft polymerization (RIGP) to synthesize adsorbents on high-surface area trunk fibers and the other method utilized a chemical grafting technique, atom-transfer radical polymerization (ATRP). Both approaches have shown promising uranium extraction capacities: RIGP adsorbent achieved 5.00 ± 0.15 g U/kg-ads., while ATRP adsorbent achieved 6.56 ± 0.33 g U/kg-ads., after 56 days of seawater exposure. These achieved values are the highest adsorption capacities ever reported for uranium extraction from seawater. The study successfully demonstrated new fiber materials with sorption capacity at 5.0 g-U/kg adsorbent under marine testing conditions. Further optimization, investigation of other new materials as well as deepening our understanding will develop adsorbents that have even higher uranium adsorption capacity, increased selectivity, and faster kinetics.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Modeling and Analysis of a One-Third Scale Used Nuclear Fuel Package 30 cm Drop

This paper compares finite element models of used nuclear fuel package drop tests to actual test data recorded in a recent test campaign. In December of 2018, the US Department of Energy collaborated with Equipos Nucleares S.A, S.M.E (ENSA) and Bundesanstalt für Materialforschung und -prüfung (BAM) to perform 30 cm horizontal drop tests of a one-third scale model of the ENSA ENUN 32P dual-purpose used nuclear fuel package. The third-scale test package included a third-scale basket and simulated third-scale fuel assemblies that were instrumented with accelerometers to measure their individual impact responses. Two horizontal package drop tests were performed, one with the basket cells oriented parallel and perpendicular to the impact surface, and one with the basket rotated 45 degrees to change the relative orientation with the impact surface. This paper describes the modeling and analysis that supported the development of the test and instrumentation plan, such as which fuel assemblies should be instrumented and at what locations on each fuel assembly should the accelerometers be placed. Pre-test predictions based on finite element modelling are discussed, and a comparison against the actual test data is made. Post-test analysis and model refinement are also discussed, with the goal of identifying best practices and lessons learned. A separate paper by Kalinina et al., to be presented at this conference, more fully describes the test configuration and test results. This paper is focused on the numerical modeling that supported the test and the lessons learned about realistically modeling this type of physical test.

Klymyshyn, Nicholas A.↗
DOE OSTI2017

Statistical Analysis in Support of Fuel Temperature Control in the Advanced Gas Reactor Experiments

There are four completed tests among a series of Advanced Gas Reactor (AGR) irradiation experiments that are being conducted within the Advanced Reactor Technologies (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaigns are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and hypothetical accident conditions, and support development and validation of fuel performance and fission product transport models and codes. Since this tri-structural isotropic fuel is qualified to be used in high temperature gas-cooled reactors (HTGRs), it is essential that the testing conditions of fuel irradiations are designed to mimic those in the conceptual gas reactors. As a result, temperatures of the AGR test fuel have to be maintained within a specified range that is typical of the HTGRs. In each AGR capsule, the test fuel temperature is controlled independently using a variable fraction neon and helium gas mixture that fills the gap volume for each capsule. The temperature as measured by a thermocouple (TC) terminated in the graphite holder (which houses fuel compacts) is used as a feedback for the gas flow controllers. Direct temperature measurement for fuel compacts is not possible because contact between a TC head and fuel particle could lead to undesirable failures. Therefore, an ABAQUS-based finite element thermal model was created to predict the daily average fuel temperatures during the entire irradiation period. It is well-known that extended exposure to neutron irradiation and high temperatures leads to changes in geometries of fuel compacts and graphite holder, which results in changes of the gas gap, leading to a change in the relationship between TC and fuel temperatures. In addition, TCs are expected to drift and sometimes fail under irradiation conditions. These facts make maintaining test fuel temperature within a specified range challenging. This paper describes how the regression analysis methods are used in conjunction with calculated fuel temperatures from early irradiation periods (when the irradiation effect is relatively small) to effectively control fuel temperature for the remaining irradiation (when the irradiation effect became profound). This is particularly important in the event when all TCs in a capsule fail.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Post-Transient Examination Results of RIA Commissioning Teats at the Transient Reactor Test Facility

Six reactivity-initiated-accident (RIA) commissioning tests have been performed at Idaho National Laboratory’s (INL) Transient Reactor Test Facility. Five of these tests were performed using fresh fuel rodlets, and the sixth test was performed using a previously irradiated rodlet from the ATF-2 irradiation experiment. These experiments demonstrate the ability to perform RIA testing including the ability to perform experiments using previously irradiated materials. Post-transient examinations revealed pellet cladding interactions and cladding ballooning depending on the initial boundary conditions of the test. Increased thickness of both zirconium oxide and alpha-zirconium were found in the cladding with tests performed at higher levels of total energy deposition. No oxide or alpha-zirconium was found in the test with the greatest amount of ballooning, which was also the test with the highest initial rod internal pressure. The highest energy deposition test (1110 J/g) resulted in failure of the rodlet. Fragmentation of the fuel pellets occurred in this test with the size of the fragments decreasing inversely to the pellet radius.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1995

Hybrid fuel formulation and technology development

The objective was to develop an improved hybrid fuel with higher regression rate, a regression rate expression exponent close to 0.5, lower cost, and higher density. The approach was to formulate candidate fuels based on promising concepts, perform thermomechanical analyses to select the most promising candidates, develop laboratory processes to fabricate fuel grains as needed, fabricate fuel grains and test in a small lab-scale motor, select the best candidate, and then scale up and validate performance in a 2500 lbf scale, 11-inch diameter motor. The characteristics of a high performance fuel have been verified in 11-inch motor testing. The advanced fuel exhibits a 15% increase in density over an all hydrocarbon formulation accompanied by a 50% increase in regression rate, which when multiplied by the increase in density yields a 70% increase in fuel mass flow rate; has a significantly lower oxidizer-to-fuel (O/F) ratio requirement at 1.5; has a significantly decreased axial regression rate variation making for more uniform propellant flow throughout motor operation; is very clean burning; extinguishes cleanly and quickly; and burns with a high combustion efficiency.

Dean, D. L.↗