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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 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Improved Calculation of the Young’s Modulus of Rectangular Prisms from Their Resonant Frequency Overtones by Identifying Appropriate Shear Constants

Young’s modulus is an important parameter for characterizing the strength of, and wave propagation through, a given material. This study improves the estimation of Young’s modulus using the impulse excitation technique based on an experimental analysis of 19 borosilicate glass bars. Analysis of the frequency equations relating Young’s modulus to the out-of-plane and in-plane flexural resonant frequencies of rectangular prisms has been conducted for both the fundamental frequency and its overtones at higher orders of vibration. Furthermore, the Young’s moduli of three novaculite rocks with various porosities were then measured up to the seventh order of vibration to validate the optimum shear constant equation for estimating Young’s modulus. Young’s modulus was found to be nearly frequency independent for these rock samples.

Impulse excitation↗
DOE OSTI2026

Calculating beam extinction in a pulsed proton beam using FPGA-based peak detection

The Mu2e experiment at Fermilab imposes stringent requirements on the elimination of out-of-time beam in its pulsed proton beam, a requirement known as “extinction”. Utilizing a new μTCA-based FPGA data acquisition system, we recorded live particle data from scattered particles incident on an array of quartz Cherenkov radiators and photomultiplier tubes to measure the extinction in the inter-pulse gaps in the pulsed proton beam. Minuscule errors in the derived signal period can make a measurement of the extinction impossible, so after taking a Fourier transform, further optimizations on the period were done based on the assumption that the signal period is stable over the full time of the beam spill while it is being resonantly extracted. After these optimizations, the beam extinction was shown to be on the level of 10^3.

Hensley, Ryan [UC, Davis]↗
DOE OSTI2024

Calculated thermal expansion of γ- and δ-Pu [Poster]

Excluding the spin-orbit coupling (SOC) for the p states, as proposed by Sadigh et al. (Applied Sciences 9, 5020 (2019)) to improve the volume, worsens the agreement with experiment of the relative energies and the thermal expansion.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Zirconium Sludge Criticality Calculations in Large Process Tanks

The Savannah River Site’s H-Canyon facility has been tasked with accelerating the disposition of used research reactor fuels. • Future missions will involve dissolution of “Non-Aluminum” clad fuels – not easily dissolved – Nitric acid is not capable of dissolving stainless steel, Zirconium, and Hastelloy, clad fuels by itself • The electrolytic dissolver has been (re) selected as a disposition path for these fuels. 2

Devine, Nathan P.↗
DOE OSTI2024

Calculation of Velocities from Explosive Shot Test Fast-Frame Imagery

A cylinder of the experimental HE is detonated inside of an aquarium and sequential fast-frame images are taken of the resulting shock wave. We wish to use the sequence of images to extract quantitative data regarding the detonation velocity and velocity of the shock wave which can then be used to tune the parameters of the equation of state for the experimental HE material.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2024

UO 2 -Liquid Metal Suspension Fuel Concept: Alloy Selection and Depletion Calculations

A novel nuclear fuel was recently proposed that consists of UO 2 particles suspended in a liquid metal mixture. Two liquid metal candidates were identified: BiPbSn and PbSn. Scoping simulations were performed on an over-simplified reactor core that showed a potential limit on achievable burnup. Those findings led to two questions that needed to be addressed: What is the corrosion behavior of the liquid metals toward Zircaloy-4 (Zr4)? And what level of burnup can the fuel be expected to achieve? This report documents efforts to answer both of those questions. First, corrosion experiments were performed between the liquid metal candidates and Zr4 coupons. Second, the previous scoping simulations were expanded to include depletion analyses so that maximum burnups could be estimated. At 400°C, neither PbSn nor BiPbSn caused significant corrosion to the Zr4 coupons. At 600°C, BiPbSn did not cause significant corrosion, but PbSn did. In addition, pure bismuth and pure tin were both highly corrosive, whereas pure lead was non-corrosive. The depletion simulations found that low-enriched fuel could maintain light-water reactor operating temperatures but achieve a burnup of only about one-third of existing reactors. Higher enrichments could achieve burnups equal to or higher than those of existing reactors but at the expense of higher operating temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Monte Carlo Simulation with CAD Interface for Calculation of 3D Maps of Residual Dose (CRADA)

Objective: To develop an easy-to-use software application to predict and mitigate radiation effects in research environment, space instruments, nuclear plants and medical facilities and help nonproliferation and national security efforts. Tech-X will develop standalone software libraries and command-line tools for ( 1) translating CAD into tessellated surfaces and tetrahedral meshes in GDML (for Geant4 and MARS 15), ROOT (for MARS 15) and HDF5 (for compact representation and for the visualization) formats, (2) healing CAD geometries to make them suitable for Monte Carlo simulations; (3) creating uniform and variable Cartesian and cylindrical meshes for detailed scoring; and ( 4) efficient Monte Carlo navigation in CAD geometries. JLAB will finish automation of simulations of residual dose in CAD geometries and integrate Tech-X software into Geant4 and MARS15. Finally, Tech-X will develop a Graphical User Interface to set up and heal CAD geometries, create input files, run and visualize simulations for residual dose. This application will run on local desktops, local and remote clusters and supercomputers and will be made available through public clouds, such as Amazon Web Services.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Complete Development of Critical Capabilities for TRISO Fission Product Source Term Calculations and Quantify Mechanisms for Pd Penetration of SiC

Overall fission product (FP) release will be an important consideration for the licensing and deployment of advanced reactors utilizing tristructural isotropic (TRISO) fuels. This work focuses on enhancing and applying the BISON models needed to predict FP transport within TRISO particles and particle failure probability, both of which factor directly into release predictions. Specifically, this report details (1) the development of the models needed to predict palladium (Pd) conservation at the engineering scale and the application of those models to characterize Pd fluxes for input into a mechanistic multiscale model for Pd penetration; (2) the refinement of sorption mass transfer models and the development of models for trapping in porous layers, which were applied and compared to particle scans from AGR-2 to provide proof of concept for a method of particle-scale validation that may reduce uncertainties compared to compact-scale validation using data from integral effects tests; (3) the development of a failure-statistics-informed, mesh-independent methodology for applying smeared cracking, enabling further study of the localized multiphysics behaviors associated with cascading particle failure mechanisms; and (4) the preliminary characterization of those coupled multiphysics particle failure behaviors using smeared, nonretentive diffusivities to provide a baseline for future study and to guide ongoing engineering applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Current State of Neutron Source Term Calculations

SOURCES-4C has some significant deficiencies that are mostly due to problems with the default data libraries that it uses. Some research groups have modified these libraries to fix most of the problems, but not all of them. Adding modern data libraries to SOURCES-4C is one of the most promising approaches to resolving these issues. Other promising options include using either NEDIS or a GEANT4 tool. Both of these codes approximately match the accuracy of the modified SOURCES-4C code. The NEDIS code is Russian. The GEANT4 tool might be computationally slow and it might also no longer be actively maintained. There are a handful of other (α,n) codes, but they are not accurate.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Fuel Performance Calculations of Tagged UO 2 MiniFuel Disks

Isotopic taggants are being considered as a novel form of forensic identifiers for nuclear fuels. As part of this effort, eighteen tagged MiniFuel disks were irradiated in the High Flux Isotope Reactor at Oak Ridge National Laboratory. In this work, the BISON fuel performance code was used to recreate the pellets and predict their behavior and their post-irradiation material properties. These predictions are to serve as a baseline to help investigators identify results of interest during post-irradiation examination of actual MiniFuel disks. The simulation methodology and models are described herein, along with the individual disk predictions. Several general trends were identified in this analysis. Based on the models, it is predicted that the pellets’ internal temperatures will rise over time, fission gas release will remain constant at around 1%, regardless of the burnup, and no grain growth will occur in any of the disks. The reasons for these predictions are discussed below. Probable shortcomings in the BISON predictions are addressed, and future work is proposed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗