Engineering Commons^™

Development Of A Detailed And Reduced-Order Neutronics Models For Fusion Reactor Blanket And Systems Design Optimization, Felipe Santos Novais

Doctoral Dissertations

Fusion neutronics plays a critical role in the design and operation of fusion reactors. However, there are a number of challenges associated with fusion neutronics that must be overcome in order to help achieve practical and sustainable fusion energy. The primary objective of this project is to develop and demonstrate a broad neutronics analysis capability in support of system studies for U.S. concepts such as the Fusion Nuclear Science Facility (FNSF). According to a detailed TBR analysis that included an examination of reaction rates and the impact of facility (FNSf) components on neutron spectra and TBR, Li-6 will generate the …

Neutron And Photon Imaging Capabilities Of Bismuth-Loaded Plastic, Andrew W. Decker

Doctoral Dissertations

Plastic scintillators utilizing iridium complex fluorophores offer substantial improvements in light yield, and their light yield is not significantly quenched in compositions with bismuth metalorganic loading at 21% weight. These advances may resolve significant capability gaps for low-cost, portable, and durable dual-particle imaging (DPI) systems for nuclear safety, security, and safeguard purposes. However, all candidate materials should first undergo investigation utilizing industry standards to quantify and evaluate their capabilities. As such, a 21% bismuth-loaded polyvinyl toluene (BiPVT) scintillator fabricated by Lawrence Livermore National Laboratory (LLNL) is computationally and experimentally evaluated as a small, pixelated radiographic array, with individual pixel dimensions …

Modeling A Nuclear Research Reactor And Radiation Dose Estimation In An Accident Scenario, Abdulaleem Abdulmajeed Bugis

Doctoral Dissertations

“A detailed, flexible three-dimensional (3D) model of the Missouri S&T Reactor (MSTR) with a heterogeneous core geometry was developed using the Standardized Computer Analyses for Licensing Evaluation (SCALE). A Graphical User Interface (GUI) was developed for SCALE, which allows the user to generate an input file automatically. The SCALE model was validated with a Monte Carlo N-Particle Transport (MCNP) model of the MSTR. The validation process was based on the criticality calculations using the Criticality Safety Analysis Sequence (CSAS6). Three geometrical models were examined. The SCALE model that has the full detailed geometry showed a good agreement with the MCNP …

Development Of A Switchable Radioisotope Generator, Kyle Mitchell Paaren

Doctoral Dissertations

The Switchable Radioisotope Generator utilizes alpha-induced reactions to produce a combination of photons, neutrons, and protons with varying fluxes dependent on target materials and source geometry. The activity/strength of the secondary radiation is further controlled by manipulating the number of alpha particles that can interact with the target material(s). Analytical equations were solved to confirm secondary radiation production from target materials using average cross sections from TENDL data. TENDL and JENDL data was confirmed by analytically solving for the total alpha-induced cross sections. This information was used to produce the provisional and utility Patent No: US20190013109A1. TENDL data was then …

Thorium-Based Mixed Oxide Fuel In A Pressurized Water Reactor: A Feasibility Analysis With Mcnp, Lucas Powelson Tucker

Doctoral Dissertations

"This dissertation investigates techniques for spent fuel monitoring, and assesses the feasibility of using a thorium-based mixed oxide fuel in a conventional pressurized water reactor for plutonium disposition. Both non-paralyzing and paralyzing dead-time calculations were performed for the Portable Spectroscopic Fast Neutron Probe (N-Probe), which can be used for spent fuel interrogation. Also, a Canberra ³He neutron detector's dead-time was estimated using a combination of subcritical assembly measurements and MCNP simulations. Next, a multitude of fission products were identified as candidates for burnup and spent fuel analysis of irradiated mixed oxide fuel. The best isotopes for these applications were …

Hybrid K-Edge Densitometry As A Method For Materials Accountancy Measurements In Pyrochemical Reprocessing, Matthew Tyler Cook

Doctoral Dissertations

Pyrochemical reprocessing is a proven method to recover useful fissile material from spent nuclear fuel. The process requires high temperatures and an inert atmosphere thus complicating the prospect of making materials accountancy measurements. Development of a measurement method for materials accountancy measurements has become necessary since pyroprocessing is receiving more attention as a possible compliment to aqueous reprocessing methods. If pyroprocessing is to be adapted from the engineering scale to a commercially viable reprocessing method a comprehensive safeguards measurement method and strategy must be developed.

Hybrid k-edge densitometry (HKED) has been applied to aqueous reprocessing measurements in commercial facilities. This …

Spatially-Dependent Reactor Kinetics And Supporting Physics Validation Studies At The High Flux Isotope Reactor, David Chandler

Doctoral Dissertations

The computational ability to accurately predict the dynamic behavior of a nuclear reactor core in response to reactivity-induced perturbations is an important subject in the field of reactor physics. Space-time and point kinetics methodologies were developed for the purpose of studying the transient-induced behavior of the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor’s (HFIR) compact core. The space-time simulations employed the three-group neutron diffusion equations, which were solved via the COMSOL partial differential equation coefficient application mode. The point kinetics equations were solved with the PARET code and the COMSOL ordinary differential equation application mode. The basic nuclear …