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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 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Phosphate tolerance of nitrogen-coordinated-iron-carbon (FeNC) catalysts for oxygen reduction reaction: A size-related hindrance effect

The greatest challenges facing the large – scale commercialization of the intermediate temperature phosphoric acid fuel cell (PAFC) technology are: (i) the high cost of Pt catalyst that is required to overcome the slow kinetics of oxygen reduction reaction (ORR) on the PAFC cathode, and (ii) deactivation of the Pt catalyst by exposure to the phosphoric acid electrolyte. Inexpensive carbon-based materials, particularly iron- spectroscopy is used to explore the phenomenon of phosphate poisoning nitrogen-coordinated carbon supported catalysts (FeNC), are promising cathodes for PAFCs. The higher ORR activation energy and phosphate poisoning resistance of FeNC present a clear advantage over the Pt-based commercial catalysts that are easily susceptible to deactivation after exposure to phosphate anions. While the immunity of FeNC catalysts to phosphate anion poisoning has been reported previously, this work explores the counter-intuitive and conditional nature of this poisoning resistance based on the properties of the carbon support used in the preparation of the FeNC catalyst. Additionally, a combination of electrochemical measurements with characterization experiments performed using X-ray photoelectron spectroscopy, infrared spectroscopy and in-situ X-ray absorption on FeNC catalysts synthesized from two different carbon supports. Results from these studies reveal a “size-related hindrance effect” wherein the ORR active Fe centers in FeNC remain inaccessible to larger anions such as phosphate when the Fe sites are present deeper inside the pores of a carbon support with high porosity. However, FeNC catalyst is poisoned by phosphate anions when synthesized using a carbon support with low porosity. This work helps gain insights into the nature of active sites in FeNC catalysts that will be useful in designing precious-metal free ORR catalysts for intermediate temperature fuel cell technologies.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
NASA NTRS2015

Initial Operation of the Nuclear Thermal Rocket Element Environmental Simulator

The Nuclear Thermal Rocket Element Environmental Simulator (NTREES) facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The NTREES facility has recently been upgraded such that the power capabilities of the facility have been increased significantly. At its present 1.2 MW power level, more prototypical fuel element temperatures nay now be reached. The new 1.2 MW induction heater consists of three physical units consisting of a transformer, rectifier, and inverter. This multiunit arrangement facilitated increasing the flexibility of the induction heater by more easily allowing variable frequency operation. Frequency ranges between 20 and 60 kHz can accommodated in the new induction heater allowing more representative power distributions to be generated within the test elements. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during testing In this new higher power configuration, NTREES will be capable of testing fuel elements and fuel materials at near-prototypic power densities. As checkout testing progressed and as higher power levels were achieved, several design deficiencies were discovered and fixed. Most of these design deficiencies were related to stray RF energy causing various components to encounter unexpected heating. Copper shielding around these components largely eliminated these problems. Other problems encountered involved unexpected movement in the coil due to electromagnetic forces and electrical arcing between the coil and a dummy test article. The coil movement and arcing which were encountered during the checkout testing effectively destroyed the induction coil in use at the time and resulted in NTREES being out of commission for a couple of months while a new stronger coil was procured. The new coil includes several additional pieces of support structure to prevent coil movement in the future. In addition, new insulating test article support components have been fabricated to prevent unexpected arcing to the test articles. Additional activities are also now underway to address ways in which the radial temperature profiles across test articles may be controlled such that they are more prototypical of what they would encounter in an operating nuclear engine. The causes of the temperature distribution problem are twofold. First, the fuel element test article is isolated in NTREES as opposed to being in the midst of many other mostly identical fuel elements in a nuclear engine. As a result, the fuel element heat flux boundary conditions in NTREES are far from adiabatic as would normally be the case in a reactor. Second, induction heating skews the power distribution such that power is preferentially deposited near the outside of the fuel element. Nuclear heating, conversely, deposits its power much more uniformly throughout the fuel element. Current studies are now looking at various schemes to adjust the amount of thermal radiation emitted from the fuel element surface so as to essentially vary the thermal boundary conditions on the test article. It is hoped that by properly adjusting the thermal boundary conditions on the fuel element test article, it may be possible to substantially correct for the inappropriate radial power distributions resulting from the induction heating so as to yield a more nearly correct temperature distribution throughout the fuel element.

Emrich, William J., Jr.↗
DOE OSTI2024

Characterization of Biomass-Based Fuels and Fuel Blends for Low-Emissions, Advanced Compression Ignition Engines (Final Report)

The goal of the Co-Optima program was to accelerate deployment of co-optimized fuels and engines that will reduce fuel consumption, and criterion pollutants and greenhouse gas (GHG) emissions for both spark and compression-ignited engines. In this project, combustion measurements from a constant pressure flow rig (CPFR) with optical access were used to assess fundamental combustion behavior of number of different candidate fuels and fuel blends. This chamber and its peripheral systems allowed stable control of ambient air pressure and temperature and fuel injection pressure and timing for hundreds of repeat injections allowing statistical interpretation of the results. The optical diagnostics enabled capturing of liquid length, vapor length, ignition timing, lift-off length as well as localized (temporal and spatially resolved) mixture refractive index, flame structure, and soot mass. The operating conditions allowed the study of combustion conditions promoting lean lifted flame combustion (LLFC), a mixing-controlled low-temperature combustion (LTC) strategy to reduce soot production in Advanced Compression Ignition (ACI) engines.

09 BIOMASS FUELS↗
DOE OSTI2020

Initial Evaluation of Microreactor Disposition Options

The United States Department of Energy is supporting the U.S advanced reactor industry through funding, legislation and regulatory development to actively pursue several microreactor design concepts. The Idaho National Laboratory (INL) is strategically positioned to support demonstration of microreactor technology in the next three to five years. This report provides and initial evaluation of the disposition options for microreactor spent nuclear fuel (SNF) generated as part of the microreactor technology demonstration program. Currently available information constitutes the basis for the options identified and discussions thereof. In the absence of detailed microreactor design information, assumptions were made to facilitate the identification of disposition options Particulars pertaining to any component of an identified disposition pathway are naturally highly microreactor design specific and in general such details are not provided. The diverse nature of potential microreactor SNF is reflected in the diverse nature of DOE owned SNF stored at INL. Therefore, it is anticipated that DOE currently stores and manages fuels that can serve as analogs for most microreactor fuel concepts. As such disposition options for microreactor SNF are expected to be much the same as that for these existing fuels. Two generic microreactor concepts have been selected for the purposes of this options assessment. The selected reactor concepts are a tristructural isotropic (TRISO) fueled high temperature gas reactor concept and a sodium/potassium bonded heat pipe reactor with uranium oxide fuel. Both concepts are assumed to be using high-assay low enriched uranium (HALEU) as the initial fuel composition. Interim storage, treatment and neutralization, material recovery, packaging and extended dry storage options are identified these reactors. The disposition options include existing INL facilities and capabilities and new facilities and capabilities developed as part of the microreactor program or as part of DOE's overall strategy for the eventual transfer of all SNF at INL to a permanent repository.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Ceramic Composite Inert Matrix Fuel Forms in High-Temperature Gas-Cooled Microreactors

Here, this work optimizes micro-prismatic high-temperature gas reactor (HTGR) designs to reduce the energy-normalized mass of spent nuclear fuel (SNF) and high-level waste (HLW) produced. The optimization was performed for the current graphite moderator and an inert matrix fuel (IMF) concept employing different composite moderators in a prismatic design architecture. The fuel matrix is magnesium oxide (MgO) with entrained tristructural-isotropic (TRISO) fuel. The moderator materials, including beryllium oxide (MgO-BeO) and beryllium (MgO-Be) at 40 vol % loading and yttrium hydride (MgO-YH x=1.9 ) and zirconium hydride (MgO-ZrH x=1.9 ) at 15 vol % loading, were entrained within the MgO host matrix. A generic graphite micro-prismatic HTGR is used as the baseline point design where the external dimensions are held constant. The composite moderator designs use 19.9% enriched uranium nitride TRISO fuel and hexagonal assemblies. For each IMF concept, an optimization study was performed to maximize the discharge burnup of the fuel by varying the TRISO packing fraction and the lattice pitch of the assemblies. The mass of SNF and HLW, other waste metrics, fuel cost, environmental impact metrics, and the activity of the SNF and HLW at 100 years and 100 000 years were calculated for the optimized IMF and graphite reference designs. The IMF results were subsequently compared to those of the graphite reference and the values for a light water reactor (LWR) and a small modular LWR. For the SNF and HLW, all the IMF concepts and the graphite reference produced less waste compared to the traditional LWR designs. However, the IMF concepts outperformed the graphite reference regarding the mass of SNF and HLW. For the other waste metrics, the IMF concepts showed reductions in fuel cost with improved environmental metrics relative to the graphite reference. Overall, the IMF concepts significantly reduced the SNF and HLW produced per unit of energy generated compared to traditional LWR designs.

TRISO↗
NASA NTRS1985

Two dimensional, transient catalytic combustion of CO-air on platinum

The light off transient of catalytic combustion of lean CO-air mixtures in a platinum coated channel of a honeycomb monolith is studied with a model that resolves transient radial and axial gradients in both the gas and the solid. For the conditions studied it is concluded that: the initial heat release occurs near the entrance at the gas-solid interface and is controlled by heterogeneous reactions; large spatial and temporal temperature gradients occur in the solid near the entrance controlled mostly by the availability of fuel; the temperature of the solid near the entrance achieves almost its steady state value before significant heating of the back; heterogeneous reactions and the gas heated up front and flowing downstream heat the back of the solid; the overall transient time is controlled by the thermal inertia of the solid and by forced convection; radiation significantly influences both transient and steady state particularly near the entrance; the oxidation of CO occurs mostly on the catalyst and becomes diffusion controlled soon into the transient.

Sinha, N.↗
DOE OSTI2022

The Feasibility and the Benefits of the Advanced Nuclear Fuel Pellet Designs with Radially Varying Fuel Zoning and Burnable Poison Concentration (Final Summary Report)

As part of the work supported by the US Department of Energy (DOE) Office of Nuclear Energy (NE) Gateway for Accelerated Innovation in Nuclear (GAIN) FY 2021 Voucher, Exelon Generation, now Constellation, and Oak Ridge National Laboratory (ORNL) entered into a cooperative research and development agreement (CRADA) to evaluate and assess the feasibility and impact of various conceptual advanced nuclear fuel pellet designs (ANFPDs). The objective of this project was to perform modeling and simulation and analyses using the advanced modeling and simulation capabilities of VERA/BISON, developed by DOE, to determine the viability and benefits of numerous advanced nuclear fuel pellet design concepts in terms of fuel cycle costs, operational safety, and margin improvement. Whereas the detailed coupled neutronic and thermal hydraulic analyses performed using VERA provided in-depth knowledge in terms of fuel cycle performance, the detailed VERA results were used in subsequent fuel performance analyses using BISON. These subsequent analyses focused on several key fuel performance criteria, such as peak fuel centerline temperature (FCT), fission gas release (FGR), gap closure, plenum pressure, and cladding hoop stress. These key fuel performance criteria were analyzed for some of the conceptual fuel designs and were compared with the results obtained for UO 2 fuel. The results provided detailed information to enable better understanding of the performance of the fuel types analyzed. Understanding the advantage of loading these conceptual fuel designs into the core is important not only to Constellation but also to the entire light-water reactor (LWR) fleet in the United States.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1990

Convective Evaporation Of Clusters Of Drops

Report presents results of continuing theoretical research in behaviors of sprayed liquid fuels at temperatures characteristic of furnaces. Two earlier papers arising from this investigation described in "Evaporation of Dense Fuel Sprays" (NPO-16954).

Bellan, Josette↗
NASA NTRS1988

Antimisting kerosene: Low temperature degradation and blending

The inline filtration characteristics of freshly blended and degraded antimisting fuels (AMK) at low temperature are examined. A needle valve degrader was modified to include partial recirculation of degraded fuel and heat addition in the bypass loop. A pressure drop across the needle valve of up to 4,000 psi was used. The pressure drop across a 325 mesh filter screen placed inline with the degrader and directly downstream of the needle valve was measured as a function of time for different values of pressure drop across the needle valve. A volume flux of 1 gpm/sq in was employed based on the frontal area of the screen. It was found that, at ambient temperatures, freshly blended AMK fuel could be degraded using a single pass degradation at 4,000 psi pressure drop across the needle valve to give acceptable filterability performance. At fuel temperatures below -20 C, degradation becomes increasingly difficult and a single pass technique results in unacceptable filtration performance. Recirculation of a fraction of the degraded fuel and heat addition in the bypass loop improved low temperature degradation performance. The problem is addressed of blending the AMK additive with Jet A at various base fuel temperatures.

Yavrouian, A.↗
DOE OSTI2021

Software Quality Assurance for EBR-II Fuels Irradiation and Physics Database (FIPD)

The Fuels Irradiation and Physics Database (FIPD) is an ongoing DOE project on archival of the EBR-II metal-alloy fuel irradiation experiments. As part of its use in support of license applications, the Quality Assurance Program Plan (QAPP) was drafted and endorsed by NRC in an effort to demonstrate its compliance with regulatory expectations. Software Quality Assurance (SQA) for the physics portion of FIPD is intended to qualify the calculated quantities such as fuel and cladding temperatures, neutron fluence and axially varying burnup estimates for irradiated fuel elements. This report covers the initial evaluation of SQA status of three neutron physics and thermo-fluid codes (REBUS, RCT and SE2RCT) that form the basis of calculated quantities for as-irradiated characteristics of the tested metallic fuel elements. The report also introduces an SQA plan to address the identified deficiencies. The REBUS, RCT, and SE2RCT codes are all part of the Argonne Reactor Code (ARC) code system. There is considerable knowledge and experience on REBUS and RCT but relatively less on SE2RCT. During FY2021, efforts focused on an assessment of how the data in the EBR-II Physics and Analysis DataBase (PADB) is generated with SE2RCT and used in FIPD. Additional tasks included considerations of uncertainties for power estimates in REBUS and RCT calculations and their impact on the combined RCT methodology. The RCT software usage in FIPD was assessed this year and the input/output details studied. A “requirements” document was created that identifies the key features of the RCT software being used in FIPD that need to have SQA documentation. A brief discussion on the history of RCT and its input is included in this report along with the basic SQA roadmap laid out in the requirements document. The SE2RCT software usage in FIPD is still being studied noting that there is no current manual. As part of the work done this year, two bugs were identified in the SE2RCT software which have a minor impact on the accuracy of the results it produces. No requirements document has been created, but one identified feature of SE2RCT being used that needs verification was its fuel pin temperature calculation. The work completed this year confirms that the approximations which will be included in the software verification report for SE2RCT are accurate. In addition to software quality assurance work for RCT and SE2RCT, an automated verification framework is proposed to simplify the software quality assurance process. The purpose of this framework is to streamline code verification and documentation while minimizing repetitive tasks for code developers and reviewers. The reduction of repeated input (between reference solution, software, and documentation input) throughout the SQA process reduces potential for human errors during the preparation of the supporting software quality records. The automation of the verification and documentation process proposed for this project leverages the existing verification structure already in place for the SAS4A/SASSYS-1 code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Effect of Nanoscale Ce 0.8 Gd 0.2 O 2–δ Infiltrant and Steam Content on Ni–(Y 2 O 3 ) 0.08 (ZrO 2 ) 0.92 Fuel Electrode Degradation during High-Temperature Electrolysis

Studies of Ni–yttria-stabilized zirconia (YSZ) fuel electrode degradation mechanisms in solid oxide electrolysis cells (SOECs) are complicated by the different possible Ni–YSZ microstructures and compositions, and the variations in the H 2 /H 2 O ratio encountered in an electrolysis stack. Here, in this study, we describe a life testing scheme aimed at providing survey results on degradation as a function of the H 2 O–H 2 composition, with life tests carried out at five different steam contents from 90% to 10%. A Ni–YSZ-supported symmetric cell geometry is employed both with and without infiltrated nanoscale gadolinia-doped ceria (GDC). Impedance spectroscopy is utilized to observe changes in electrochemical characteristics during the life test, and a transmission-line-based equivalent circuit is used to model the data. Post-test electrode microstructures were observed. The results suggest that the GDC infiltrant reduces the electrode polarization resistance and provides more stable electrode polarization over a range of conditions.

08 HYDROGEN↗
NASA NTRS1993

Nuclear gas core propulsion research program

Viewgraphs on the nuclear gas core propulsion research program are presented. The objectives of this research are to develop models and experiments, systems, and fuel elements for advanced nuclear thermal propulsion rockets. The fuel elements under investigation are suitable for gas/vapor and multiphase fuel reactors. Topics covered include advanced nuclear propulsion studies, nuclear vapor thermal rocket (NVTR) studies, and ultrahigh temperature nuclear fuels and materials studies.

Diaz, Nils J.↗
DOE OSTI2026

Cummins PEM Fuel Cell System for Heavy Duty Applications

The motivation for this project was to develop a high-pressure 100+ kW fuel cell stack that can be used in a modular fashion to power HD applications up to 1 MW. One of the issues that was identified during fuel cell system integration was the need for extra space for heat rejection systems(radiators) as well as additional fans and associated energy consumption to compensate for lower operating temperatures of fuel cells compared to internal combustion (IC) engines. To maximize the heat rejection capability of existing water-based coolant systems that are prevalent on IC engine-based systems, fuel cell stack design was developed to withstand and operate at IC engine coolant temperatures or higher, which in turn helps to reduce the radiator size requirement.

08 HYDROGEN↗
DOE OSTI2024

Optically multiplexed neutron time-of-flight technique for inertial confinement fusion

Neutron time-of-flight (nTOF) detectors are crucial in diagnosing the performance of inertial confinement fusion (ICF) experiments, which implode targets of deuterium–tritium fuel to achieve thermonuclear conditions. These detectors utilize the fusion neutron energy spectrum to extract key measurements, including the hotspot ion temperature and fuel areal density. Previous work [Danly et al., Rev. Sci. Instrum. 94, 043502 (2023)] has demonstrated adding 1D spatial resolution to an nTOF-like detector using a neutron aperture and streak camera to measure the ion temperature profile of an ICF implosion. By contrast, the study presented herein explores modifying the 1D detector to use a fast photomultiplier tube (PMT) to validate the design of a 2D spatially resolved instrument based on reconstruction from 1D profiles. The modification would collect time-of-flight traces from separate scintillators in an imaging array with one PMT using optical fibers of varying lengths to time-multiplex the signals. This technique has been demonstrated in ride-along experiments on the OMEGA laser with 20 fiber-coupled scintillator channels connected to a Photek PMT210. Results provide constraints on the fiber lengths and PMT gating requirements to promote pulse fidelity throughout all channels. Calibration of the detector to fixed nTOFs can provide a preliminary estimate of the instrument response function (IRF), although measurement of the IRF is currently under way. These results suggest that nTOF signals can potentially be time-multiplexed with fibers so long as the design is strategic to mitigate signal-to-noise reduction, modal dispersion, and charge build-up in the PMT, which has implications beyond ion temperature imaging.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS2001

Quantitative Surface Emissivity and Temperature Measurements of a Burning Solid Fuel Accompanied by Soot Formation

Surface radiometry is an established technique for noncontact temperature measurement of solids. We adapt this technique to the study of solid surface combustion where the solid fuel undergoes physical and chemical changes as pyrolysis proceeds, and additionally may produce soot. The physical and chemical changes alter the fuel surface emissivity, and soot contributes to the infrared signature in the same spectral band as the signal of interest. We have developed a measurement that isolates the fuel's surface emissions in the presence of soot, and determine the surface emissivity as a function of temperature. A commercially available infrared camera images the two-dimensional surface of ashless filter paper burning in concurrent flow. The camera is sensitive in the 2 to 5 gm band, but spectrally filtered to reduce the interference from hot gas phase combustion products. Results show a strong functional dependence of emissivity on temperature, attributed to the combined effects of thermal and oxidative processes. Using the measured emissivity, radiance measurements from several burning samples were corrected for the presence of soot and for changes in emissivity, to yield quantitative surface temperature measurements. Ultimately the results will be used to develop a full-field, non-contact temperature measurement that will be used in spacebased combustion investigations.

Piltch, Nancy D.↗
NASA NTRS1946

Knock-Limited Performance of Triptane and Xylidines Blended with 28-R Aviation Fuel at High Compression Ratios and Maximum-Economy Spark Setting

An investigation was conducted to evaluate the possibilities of utilizing the high-performance characteristics of triptane and xylidines blended with 28-R fuel in order to increase fuel economy by the use of high compression ratios and maximum-economy spark setting. Full-scale single-cylinder knock tests were run with 20 deg B.T.C. and maximum-economy spark settings at compression ratios of 6.9, 8.0, and 10.0, and with two inlet-air temperatures. The fuels tested consisted of triptane, four triptane and one xylidines blend with 28-R, and 28-R fuel alone. Indicated specific fuel consumption at lean mixtures was decreased approximately 17 percent at a compression ratio of 10.0 and maximum-economy spark setting, as compared to that obtained with a compression ratio of 6.9 and normal spark setting. When compression ratio was increased from 6.9 to 10.0 at an inlet-air temperature of 150 F, normal spark setting, and a fuel-air ratio of 0.065, 55-percent triptane was required with 28-R fuel to maintain the knock-limited brake power level obtained with 28-R fuel at a compression ratio of 6.9. Brake specific fuel consumption was decreased 17.5 percent at a compression ratio of 10.0 relative to that obtained at a compression ratio of 6.9. Approximately similar results were noted at an inlet-air temperature of 250 F. For concentrations up through at least 20 percent, triptane can be more efficiently used at normal than at maximum-economy spark setting to maintain a constant knock-limited power output over the range of compression ratios tested.

Held, Louis F.↗
DOE OSTI2023

Editorial: Benchmark experiments, development and needs in support of advanced reactor design

Advanced nuclear reactor designs will for the most part be a departure from low enrichment light water reactor (LWR) designs currently operated around the world. Such advanced designs include but are not limited to new TRISO-fueled high temperature gas reactors, heat-pipe cooled micro-reactors, fluoride salt cooled high-temperature reactors, molten salt reactors, lead cooled fast reactors, nuclear thermal propulsion concepts, and include LWR designs with advanced fuel and clad types. Modeling and simulation methods for advanced reactors is necessary for regulators to approve license requests. However, regulators also require that modeling approaches be validated against experimental measurements. Hence, there is a crucial need for data for advanced reactor systems that will support validation of analysis methods. To this end, this Research Topic includes eleven papers organized into topical seven categories relevant for advanced reactor design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗