Index Catalog // ScholarsArchive@OSU

Incorporating statistical uncertainty into nuclear material diversion pathway analysis using probabilistic risk assessment

Creator:: Pharn, Benjamin T.
Abstract:: A Level 1 probabilistic risk assessment (PRA) is applied to a plutonium button-making process, but in regards to the risk of diversion (or theft) of special nuclear material (SNM) rather than the risk of mechanical failure. The main objective of the project was to identify the relative vulnerability of areas...
Resource Type:: Masters Thesis
Full Text:: button- making process, but in regards to the risk of diversion (or theft) of special nuclear material

Detection of actinides via nuclear isomer de-excitation

Creator:: Francy, Christopher J.
Abstract:: The detection of illicit weapons-grade nuclear material is a significant technical challenge receiving a great deal of attention in the nuclear community. A proposed solution is to actively bombard the material under investigation with neutrons, populating nuclear isomers in any potential fissile material. These isomers decay quickly, releasing gamma rays...
Resource Type:: Masters Thesis
Full Text:: : __________________________________________________________ Todd S. Palmer The detection of illicit weapons-grade nuclear material is a significant

An assessment of a mesh-based method for slab-geometry transport in stochastic materials

Creator:: Bassett, Brody R.
Abstract:: There exists a variety of methods to solve the radiation transport equation in stochastic materials, including the Levermore-Pomraning method and the atomic mix method, each of which is designed to replicate the results given by benchmarks with materials rearranged in a randomized fashion. While the atomic mix method is very...
Resource Type:: Masters Thesis
Full Text:: equation for each material containing a term coupling it to each of the other materials [3]. This method

Applying Machine Learning Algorithms to Identify Problematic Nuclear Data within Nuclear Transport Simulations

Creator:: Grechanuk, Pavel A.
Abstract:: The primary goal of this dissertation is to improve the quality of nuclear data available to the nuclear science community. We propose to accomplish this by applying machine learning algorithms to the large number of available benchmark experiments and simulations, with the goal of determining which nuclear data have strong...
Resource Type:: Dissertation
Full Text:: sensitivity coefficients detail what materials and reactions are important to the behavior of a nuclear

Steady State Modeling of the Minimum Critical Core of the Transient Reactor Test Facility

Creator:: Alberti, Anthony L.
Abstract:: With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited...
Resource Type:: Masters Thesis
Full Text:: evaluate nuclear fuels and associated structural materials in a variety of excursion scenarios. TREAT is

Analysis of the antineutrino rate during CANDU reactor startup

Creator:: Matthews, Christopher
Abstract:: Detection systems used to monitor reactor operations are of significant interest as tools for verification of operator declarations. Current reactor site safeguards are limited to visual inspections and intrusive monitoring systems. The recent development of antineutrino detectors may soon allow real-time monitoring from an unobtrusive location. Antineutrinos are produced through...
Resource Type:: Masters Thesis
Full Text:: for the degree of Master of Science in Nuclear Engineering presented on January 27, 2012. Title

Design and analysis of a nuclear reactor core for innovative small light water reactors

Creator:: Soldatov, Alexey I.
Abstract:: In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design...
Resource Type:: Dissertation
Full Text:: 0 Department of Nuclear Engineering And

Neutronic Analysis of a Liquid Metal Fast Reactor Experiment in the Large-I Position of the Advanced Test Reactor

Creator:: Lilley, Benjamin C.
Abstract:: Material testing experiments are needed to enable the next generation of nuclear fission reactors. Concluding in 2009, the Boosted Fast Flux Loop (BFFL) project was devised as a way to test fast reactor materials and fuels in the Advanced Test Reactor (ATR), however the experiment location it used is now...
Resource Type:: Masters Thesis
Full Text:: Flux Test Facility,” Journal of Nuclear Materials, 204, 109–118 (1993). [9] C. WEMPLE, “Design of a

The iterative thermal emission Monte Carlo method for thermal radiative transfer

Creator:: Long, Alex R.
Abstract:: For over 30 years, the Implicit Monte Carlo (IMC) method has been used to solve challenging problems in thermal radiative transfer. These problems are typically optically thick and di ffusive, as a consequence of the high degree of "pseudo-scattering" introduced to model the absorption and reemission of photons from a...
Resource Type:: Masters Thesis
Full Text:: Master of Science in Nuclear Engineering presented on June 1 2012. Title: The Iterative Thermal Emission

Modeling the Oregon State University TRIGA Reactor Using the Attila Three-Dimensional Deterministic Transport Code

Creator:: Keller, S. Todd
Abstract:: Attila is a state-of-the-art radiation transport based code that is efficient, accurate and straightforward to use. Geometric information is input to the code as CAD drawings. Material property information is input as a cross section library. The user has full control over standard transport options and output reports. In phase...
Resource Type:: Masters Thesis
Full Text:: Materials Most core components are constructed as a central region of one material with a thin layer

ScholarsArchive@OSU

Incorporating statistical uncertainty into nuclear material diversion pathway analysis using probabilistic risk assessment

Detection of actinides via nuclear isomer de-excitation

An assessment of a mesh-based method for slab-geometry transport in stochastic materials

Applying Machine Learning Algorithms to Identify Problematic Nuclear Data within Nuclear Transport Simulations

Steady State Modeling of the Minimum Critical Core of the Transient Reactor Test Facility

Analysis of the antineutrino rate during CANDU reactor startup

Design and analysis of a nuclear reactor core for innovative small light water reactors

Neutronic Analysis of a Liquid Metal Fast Reactor Experiment in the Large-I Position of the Advanced Test Reactor

The iterative thermal emission Monte Carlo method for thermal radiative transfer

Modeling the Oregon State University TRIGA Reactor Using the Attila Three-Dimensional Deterministic Transport Code

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