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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 541 records · Page 30
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Performance characterization of CNTs and γ-Al{sub 2}O{sub 3} supported cobalt catalysts in Fischer-Tropsch reaction

Catalysts were prepared via a wet impregnation method. Different physicochemical properties of the samples were revealed by transmission electron microscope (TEM), temperature programmed reduction (H{sub 2}-TPR) and carbon dioxide desorption (CO{sub 2}-desorption). Fischer-Tropsch reaction (FTS) was carried out in a fixed-bed microreactor at 220°C and 1 atm, with H{sub 2}/CO = 2v/v and space velocity, SV of 12L/g.h for 5 h. Various characterization techniques revealed that there was a stronger interaction between Co and Al{sub 2}O{sub 3} support compared to that of CNTs support. CNTs support increased the reducibility and decreased Co particle size. A significant increase in % CO conversion and FTS reaction rate was observed over CNTs support compared to that of Co/Al{sub 2}O{sub 3}. Co/CNTs resulted in higher C{sub 5+} hydrocarbons selectivity compared to that of Co/Al{sub 2}O{sub 3} catalyst. CNTs are a better support for Co compared to Al{sub 2}O{sub 3}.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

Review of electric discharge microplasmas generated in highly fluctuating fluids: Characteristics and application to nanomaterials synthesis

Plasma-based fabrication of novel nanomaterials and nanostructures is indispensible for the development of next-generation electronic devices and for green energy applications. In particular, controlling the interactions between plasmas and materials interfaces, and the plasma fluctuations, is crucial for further development of plasma-based processes and bottom-up growth of nanomaterials. Electric discharge microplasmas generated in supercritical fluids represent a special class of high-pressure plasmas, where fluctuations on the molecular scale influence the discharge properties and the possible bottom-up growth of nanomaterials. This review discusses an anomaly observed for direct current microplasmas generated near the critical point, a local decrease in the breakdown voltage. This anomalous behavior is suggested to be caused by the concomitant decrease of the ionization potential due to the formation of clusters near the critical point, and the formation of extended electron mean free paths caused by the high-density fluctuation near the critical point. It is also shown that in the case of dielectric barrier microdischarges generated close to the critical point, the high-density fluctuation of the supercritical fluid persists. The final part of the review discusses the application of discharges generated in supercritical fluids to synthesis of nanomaterials, in particular, molecular diamond—so-called diamondoids—by microplasmas generated inside conventional batch-type and continuous flow microreactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Anodic stripping voltammetry of synthesized CdS nanoparticles at boron-doped diamond electrodes

Cadmium sulphide (CdS) nanoparticles were chemically synthesized using reverse micelles microreactor methods. By using different washing treatments, UV-Vis spectroscopy showed that the absorption peaks appeared at 465 nm, 462 nm, 460 nm, and 459 nm respectively for CdS nanoparticles without and with 1, 2, and 3 times washing treatments using pure water. In comparison with the absorbance peak of bulk CdS at 512 nm, the shifted absorption peaks, indicates that the different sizes of CdS can be prepared. Anodic stripping voltammetry of the CdS nanoparticles was then studied at a boron-doped diamond electrode using 0.1 M KClO{sub 4} and 0.1 M HClO{sub 4} as the electrolytes. A scan rate of 100 mV/s with a deposition potential of -1000 mV (vs. Ag/AgCl) for 60 s at a potential scan from -1600 mV to +800 mV (vs. Ag/AgCl) was applied as the optimum condition of the measurements. Highly-accurate linear calibration curves (R{sup 2} = 0.99) in 0.1 M HClO{sub 4} with the sensitivity of 0.075 mA/mM and the limit of detection of 81 µM in 0.1 M HClO{sub 4} can be achieved, which is promising for an application of CdS nanoparticles as a label for biosensors.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2019

In operando studies of CO oxidation on epitaxial SrCoO 2.5+δ thin films

The high abundance and fast kinetics of select transition metal oxide catalysts are attractive features for many chemical and electrochemical device applications. However, the activity of such catalysts can be accompanied by phase instabilities that prevent their widespread usage. Furthermore, complexities associated with variations in phase behavior and oxygen stoichiometry have hindered studies on the true origins of catalytic activity. Here, we explore the interactions between activity, phase stability, and microstructure using in operando synchrotron X-ray techniques and gas chromatography/mass spectroscopy (GCMS) to probe the behavior of model SrCoO 2.5+δ (SCO) catalysts. Pulsed laser deposition was used to prepare SCO thin films on (001) SrTiO 3 , (111) SrTiO 3 , and pseudocubic (001) DyScO 3 substrates. The GCMS catalytic measurements were performed with a custom-built microreactor compatible with a synchrotron X-ray diffractometer at the Advanced Photon Source. The activity for carbon monoxide oxidation was determined as a function of temperature from 500 °C to 800 °C. We show that the SrCoO 2.5+δ films are active for CO oxidation, most likely by direct reaction with lattice oxygen; consequently, the activity was observed to increase as the films become less stable, with the most active film being the one exhibiting the lowest surface and crystal quality. All films decompose at high temperatures, with in operando diffraction indicating the gradual formation of Sr-rich hexagonal and CoO phases. We find that real-time studies of model oxide systems with synchrotron X-rays is a powerful means of gaining insight into the varied processes taking place at catalytic surfaces.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Spatiotemporal periodicity of dislocation dynamics in a two-dimensional microfluidic crystal flowing in a tapered channel

Collective interactions in many-body systems can give rise to unexpected order. Such interactions underlie a wide range of complex phenomena such as swarming in animals and traffic patterns. This work reports an unexpected order in the flow of highly confined and highly concentrated water-in-oil drops, which can be explained and modeled as a soft crystal being extruded in the nanoscale. The findings are important to the understanding of out-of-equilibrium many-body systems, and the flow control of these drops used as microreactors in droplet microfluidics. Furthermore, contrary to the unpredictable microscale crystal deformation process, the discoveries here indicate that nanoscale crystal deformation can be highly ordered and predictable, and imply that the manufacturing of nanocrystalline materials may be easier than perceived.

42 ENGINEERING↗
DOE OSTI2020

Hanging droplets from liquid surfaces

Natural and man-made robotic systems use the interfacial tension between two fluids to support dense objects on liquid surfaces. Here, we show that coacervate-encased droplets of an aqueous polymer solution can be hung from the surface of a less dense aqueous polymer solution using surface tension. The forces acting on and the shapes of the hanging droplets can be controlled. Sacs with homogeneous and heterogeneous surfaces are hung from the surface and, by capillary forces, form well-ordered arrays. Locomotion and rotation can be achieved by embedding magnetic microparticles within the assemblies. Direct contact of the droplet with air enables in situ manipulation and compartmentalized cascading chemical reactions with selective transport. Applications including functional microreactors, motors, and biomimetic robots are evident.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Sockeye: A One-Dimensional, Two-Phase, Compressible Flow Heat Pipe Application

Sockeye is a heat pipe analysis application based on the Multiphysics Object-Oriented Simulation Environment (MOOSE) finite element framework. The primary purpose of Sockeye is to provide a transient heat pipe simulation tool to be used in the analysis of nuclear microreactor designs. Sockeye provides the capability to perform one-dimensional, two-phase, compressible flow simulation of a heat pipe working fluid and two-dimensional, axisymmetric heat conduction for the heat pipe cladding and its surroundings. Sockeye is demonstrated against analytical solutions and experimental data from the SAFE-30 heat pipe module test.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

A MOOSE-Based Model for Fission Product Transport and Source Term Estimation for High-Temperature Gas-Cooled Reactors

Thanks to fuel elements containing tristructural isotropic (TRISO) particles combined with a low core power density and passive feedback mechanisms leading to modest temperature rises in the event of accidental events, high-temperature gas-cooled reactors (HTGRs) offer a high degree of reliability in terms of fission product retention. While the anticipated source term for HTGRs is expected to be very low, it is important to provide a quantitative estimate of radiological releases during nominal and accidental conditions. Here, we propose a computationally efficient mechanistic source term methodology relying on the Multiphysics Object Oriented Simulation Environment (MOOSE) for tracking fission product transport from TRISO particles up to the coolant pressure boundary, as well as modeling the transport and potential deposition of these nuclides inside the reactor coolant loop. The proposed computational scheme is applied to estimate source term inventories for a representative 10-MW(thermal) prismatic high-temperature microreactor and is qualitatively compared against known release fractions. In addition to providing an alternate analysis tool, this MOOSE model can help reactor designers quantify the influence of key design parameters relevant for studies of radiological dose consequences.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

RELAP5-3D Capabilities for High-Temperature Gas-Cooled Reactor Analysis

RELAP5-3D is the premier systems thermal-hydraulic code in the world. The code has been used significantly for high-temperature gas-cooled reactor (HTGR) analyses over the last 10 years, but there has not been a comprehensive summary of RELAP5-3D capabilities and methods for HTGR modeling. In this paper, we provide a guide for how novice users can build models for prismatic HTGRs. This guide includes best practices for developing heat structures and setting up radiation and conduction heat transfer. We also provide a summary of relevant validation studies. These studies are divided into three categories: separate effects tests, integral effects tests, and nuclear reactor studies. We present a comprehensive review of RELAP5-3D HTGR applications for reactor modeling over the last 10 years. These applications include work for the Modular High-Temperature Gas-Cooled Reactor benchmark, design activities for a prismatic HTGR as part of the U.S. Department of Energy’s Advanced Demonstration and Test Reactor Options Study, and studies performed to design gas-cooled microreactors. Following the comprehensive literature review, we present a gap analysis highlighting capabilities of RELAP5-3D that are available and missing for a complete analysis of HTGR thermal hydraulics. We also discuss the limitations of existing code capabilities and potential improvements that could improve analyses. Current research efforts have yet to fully validate RELAP5-3D for a prismatic HTGR analysis; however, some of this is attributable to confounding factors in the available experimental data as opposed to code capabilities. In conclusion, this paper presents the most comprehensive review to date on RELAP5-3D capabilities, limitations, and applications for HTGR analysis.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Assessment of Critical Experiment Benchmark Applicability to a Large-Capacity HALEU Transportation Package Concept

This work presents an assessment of the applicability of existing benchmark critical experiments to the criticality safety code validation for a large-capacity high-assay low-enriched uranium (HALEU) transportation package concept. Numerous next-generation nuclear reactor designs require HALEU fuel, which is characterized by an enrichment between 5 and 20 wt% 235 U. The U.S. Department of Energy (DOE) has proposed to recover and downblend highly enriched uranium from DOE-owned used nuclear fuel to accelerate the demonstration of commercially viable microreactor technologies. One element of the infrastructure needed to demonstrate HALEU-fueled reactors is the ability to safely transport enriched product to be used for fuel fabrication. There is uncertainty as to whether existing critical benchmark experiment data are sufficient to support criticality safety code validation for HALEU transportation applications. The anticipated chemical form of the HALEU in the proposed transportation concept is UO 2 with 20 wt% 235 U/U. The concept uses a combination of an existing transportation packaging design and a novel basket design, including borated aluminum flux traps. The basket provides space for 18 reusable, stainless steel canisters that contain the HALEU. In 10 CFR 71, normal conditions of transport (NCTs) and hypothetical accident conditions (HACs) are defined for fissile material transportation packages. NCT and HAC KENO-VI models of the transportation package were developed using the Standardized Computer Analyses for Licensing Evaluation (SCALE) 6.2.3 computer code package, and optimum moderation conditions were determined using the SCALE SAMPLER sequence. The SCALE Tools for Sensitivity and Uncertainty Analysis Methodology Implementation (TSUNAMI) sequences were then used to compare the neutronic characteristics of 1584 International Criticality Safety Benchmark Evaluation Project benchmark critical experiments with the NCT and HAC HALEU transportation models. The TSUNAMI integral correlation coefficient c k was the criterion used to rank neutronic similarity. Thirty-four experiments were identified as similar (c k ≥ 0.9) to the NCT model, and 55 experiments were identified as similar to the HAC model. Hundreds of experiments were also identified as at least marginally similar (c k ≥ 0.8) to both models. The results indicate that additional critical experiments are unlikely to be needed to support HALEU transportation criticality safety analyses for package concepts similar to the concept package analyzed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

MOOSE Reactor Module: An Open-Source Capability for Meshing Nuclear Reactor Geometries

The U.S. Department of Energy (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has developed numerous physics solvers utilizing the open-source Multiphysics Object-Oriented Simulation Environment (MOOSE) framework for multiphysics reactor analysis. These solvers require input finite element meshes representing the discretized spatial domain. Typically, reactor analysts turn to licensed tools for the creation of reactor geometry meshes. Recently, open-source functionality has been added to the MOOSE framework to mesh common reactor geometries and improve MOOSE-based nuclear reactor application user workflows. The new functionality is primarily contained in the new Reactor module of MOOSE and includes support for hexagonal pins, assemblies, and cores, extended Cartesian geometry support, options for modeling static and rotating control drums within a hexagonal assembly, core periphery triangulation, and automatic tagging of pin, assembly, plane, and depletion regions for easier post processing of physics results. A set of reactor geometry mesh builder objects further streamlines the construction of hexagonal and Cartesian cores and allows mapping of materials to regions during mesh generation. The meshes produced with the MOOSE Reactor module may be used directly within MOOSE-based applications or exported as Exodus II files for use in other finite element solvers. The tools have been demonstrated and verified using a variety of NEAMS physics solvers on a range of reactor applications, including a sodium-cooled fast reactor core analysis using Griffin, a fast reactor assembly thermal deformation analysis using MOOSE Tensor Mechanics, and a heat pipe–cooled microreactor coupled analysis using Griffin, Bison, and Sockeye. MOOSE’s Reactor module provides significant advantages compared to the use of external meshing tools when analyzing Cartesian and hexagonal reactor lattices using MOOSE-based applications: immediate accessibility (open-source) to the end user, low barrier to entry for new users, speed of mesh generation, volume preservation of meshed fuel pins, and simplification of analysis workflow when used in conjunction with MOOSE-based applications.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2024

Optimizing Hydride Stability in U-ZrH x Nuclear Fuel: The “Goldilocks Radius”

Nuclear-powered microreactors show great promise for opening new nuclear energy markets due to the flexibility offered by their rapid/streamlined in-factory fabrication, transportability, and self-regulating nature. The economic benefits of any commercialized nuclear reactor, however, rely on the system’s ability to produce large amounts of heat and efficiently convert that heat into electrical power reliably for long periods of time. Uranium-zirconium hydride (U-ZrH x ) is currently being considered for compact reactor designs because it is a well-known nuclear fuel system that is self-moderating, but this fuel, which has historically been used for research reactors, has not been optimized for commercial power production. Here, this paper analyzes the hydride stability of standard 304 stainless steel–clad U-ZrH x fuel under commercially relevant conditions. Fuel element design parameters, including physical dimensions, as-fabricated hydrogen content, burnup, peak fuel temperature, temperature gradient, operational fuel cycle duration, and volumetric heat generation rate, are discussed with a focus on hydrogen distribution and phase stability within the fuel element. Hydride stability declines more rapidly as the coolant temperature, burnup, and fuel cycle duration increase. Using a fuel-cladding gap material with heat transfer properties superior to air, such as helium or sodium, is essential to prolonging fuel hydride stability. The fuel’s physical dimensions are also important. At very small fuel diameters, the H/Zr ratio in the fuel meat decreases too rapidly due to the hydrogen content’s dependence on fuel meat volume. Conversely, the fuel meat temperature and temperature gradient exacerbate hydrogen loss at very large fuel diameters. We find that the most important parameter to consider when optimizing the hydride stability of U-ZrH x fuel is the relationship between the fuel meat radius and the power density in the fuel. A simple equation is empirically determined that relates the “Goldilocks radius,” that is, the fuel radius for which the H/Zr ratio is most stable, to the power density in the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Welch Method and Bootstrapping Applied to Subcritical Gamma Noise

We measured the prompt neutron decay constant 𝛼 of the CROCUS zero-power reactor at the Swiss Federal Institute of Technology Lausanne using cross-power spectral density (CPSD) analysis of gamma-gamma correlations from two trans-stilbene organic scintillators positioned near the reactor core. We measured critical and subcritical states, with water levels ranging from 960 mm (critical) to 800 mm (𝜌=−1.4 $ subcritical). Our analysis used the Welch method, dividing signal segments for fast Fourier transform (FFT) frequency analysis and applying bootstrapping uncertainty quantification that uses Welch-defined segments. Results demonstrated a clear increase in the measured 𝛼 as reactor reactivity decreased, distinguishing critical from subcritical conditions. At the 960-mm critical level, 𝛼 was estimated at 155.9 ± 0.7 s −1 , and for the 800-mm subcritical level, 𝛼 increased significantly to 367.3 ± 6.9 s –1 . A linear regression of subcritical states yielded a critical estimate of 154.0 ± 3.1 s –1 , aligning with the static 𝛼 estimate at critical. The bootstrapping method produced normally distributed 𝛼 estimates, confirming data consistency. The gamma CPSD 𝛼 estimates clearly distinguish reactor states and improve monitoring of zero-power reactors. The future deployment of modular and microreactors as potential candidates for noise analysis is demonstrated in CROCUS, particularly zero-power mock-ups of new designs. The improvement of noise analysis in the subcritical domain from this work will support experimental data for reactor deployment and procedure.

CROCUS↗
DOE OSTI2025

SCALE Shielding Calculations for Advanced Reactor Accident Scenarios

The study results presented in this paper demonstrate the capabilities of the SCALE computer code for non–light water reactor (non-LWR) radiation source term and shielding calculations in support of the U.S. Nuclear Regulatory Commission non-LWR fuel cycle demonstration project. Representative non-LWR types, including the sodium-cooled fast reactor (SFR), the molten salt reactor (MSR), and the heat pipe microreactor (HPMR), were analyzed to evaluate dose rates associated with postulated accident scenarios (SFR and MSR), reactor operation (MSR and HPMR), and fuel self-protecting characteristics (HPMR). New features were implemented in SCALE depletion codes to better simulate MSR operation.

MSR↗
DOE OSTI2026

Evaluation of Radiation Transport Through a Nuclear-Grade Sandwich Composite

The development and deployment of microreactors are often synonymous with their transportability; however, designing a shield which is appropriate for both transportation and operations is challenging. A recent approach is the use of composite style shield which combines the reactor pressure vessel and biological shielding into a single unit to enhance the shields effectiveness without degrading the structural properties. To assess the viability of a nuclear grade sandwich composite (NGSC), and understanding of how radiation transports through the shield needs to be understood to select the appropriate materials. This work examines a two-layered NGSC of varying thicknesses with two primary shielding materials (tungsten-tetraboride cermet and boron carbide). To assess the NGSC, quantities of interest including neutron attenuation, photon attenuation, energy deposition, displacements per atom, helium generation, and dose are examined. Altogether, it is found that a combination of both tungsten-tetraboride and boron carbide are necessary to act as an appropriate shield, where the order in which these materials are place in the composite are important to their effectiveness.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Temperature Field Reconstruction of Surfaces Heated Through Radiative Heat Transfer Using Convolutional Neural Networks

Microreactors could play a crucial role in decarbonizing our energy portfolio. However, their development and implementation come with specific challenges, particularly regarding cost. Due to their compact size and the harsh operational environment, collecting real-time data on reactor operation can be challenging. Many probe designs are unable to withstand extreme conditions (e.g., temperature, radiation) in the reactor. In this context, using convolutional neural networks (CNNs) can pave the way for developing a nonintrusive approach that relies solely on ex-core sensors. A well-trained physics-informed CNN can reconstruct the distribution of a given physical quantity over a domain using only a few sensors, allowing us to reconstruct the desired field distribution even in a limited space or complex geometries where a large array of sensors is impractical. In this work, we present the initial steps toward developing a real-time tool for monitoring the thermal behavior of nuclear reactor pressure vessels. Based on an experimental setup, a computational model using the Multiphysics Object-Oriented Simulation Environment (moose) framework was built, where the Ray Tracing and Heat Conduction modules were used to evaluate the temperature distribution over a convex metal surface heated through radiative heat transfer. This metal surface represents a section of a heated nuclear reactor vessel wall. The model also accounts for solid mechanics physics through the moose Solid Mechanics module. In situ experimental data, acquired from a Texas A&M facility, were used to validate the computational model. Part of the data generated by the moose model was used to train the convolutional neural network to reconstruct the vessel wall's outer surface temperature. The CNN generalization was then compared against the experimental and computational data.

Aldeia Machado, Luiz Carlos↗
DOE OSTI2025

Gas-phase synthesis of anthracene and phenanthrene via radical-radical reaction induced ring expansions

Unraveling reaction mechanisms of aromatic and resonance-stabilized radicals is critical to understanding molecular mass growth processes to polycyclic aromatic hydrocarbons (PAHs) and carbonaceous nanoparticles in distinct astrophysical environments (molecular clouds, circumstellar envelopes) and combustion systems. Using photoelectron photoion coincidence spectroscopy (PEPICO), we explored the gas-phase reaction of the methyl radical (CH 3 • ) with the aromatic and resonance-stabilized fluorenyl radical (C 13 H 9 • ) under high-temperature conditions in a chemical microreactor. Anthracene and phenanthrene were detected isomer-selectively using photoionization efficiency (PIE) curves and mass-selected threshold photoelectron (ms-TPE) spectra. While phenanthrene is produced through a radical-radical recombination of the carbon-centered radicals, anthracene may plausibly be formed through an unconventional radical addition to a low spin-density fluorenyl carbon. These pathways result in five-membered ring expansion—a critical mechanism crucial to PAH mass growth converting bent PAHs into planar nanostructures.

Goettl, Shane J. [University of Hawai‘i at Mānoa, ↗
DOE OSTI2025

Comparison of nested geometry treatments within GPU-based Monte Carlo neutron transport simulations of fission reactors

Monte Carlo (MC) neutron transport provides detailed estimates of radiological quantities within fission reactors. This involves tracking individual neutrons through a computational geometry. CPU-based MC codes use multiple polymorphic tracker types with different tracking algorithms to exploit the repeated configurations of reactors, but virtual function calls have high overhead on the GPU. The Shift MC code was modified to support GPU-based tracking with three strategies: dynamic polymorphism with virtual functions, static polymorphism, and a single tracker type with tree-based acceleration. On the Frontier supercomputer these methods achieve 77.8%, 91.2%, and 83.4%, respectively, of the tracking rate obtained using a specialized tracker optimized for rectilinear-grid-based reactors. This indicates that all three methods are suitable for typical reactor problems in which tracking does not dominate runtime. The flexibility of the single tracker method is highlighted with a hexagonal-grid microreactor problem, performed without hexagonal-grid-specific tracking routines, providing a 2.19× speedup over CPU execution.

97 MATHEMATICS AND COMPUTING↗