Nuclear Engineering Commons^™

Modeling And Characterizing The Electron Backscatter In A Cylindrical Anode-Based Distributed X-Ray Source, Jordan Fox, Seth Kilby, Hyoung Koo Lee, Ayodeji Alajo, Ashish Avachat

Nuclear Engineering and Radiation Science Faculty Research & Creative Works

Upcoming advancements in computed tomography architectures warrants the investigation of new X-ray source designs and the impacts that electron backscatter can have on these designs. One such design being investigated is a distributed, cylindrical anode-based X-ray source. For such a distributed X-ray source, we developed a modeling pipeline for simulating electron optics and transport to characterize the quality of the primary X-ray beam and the electron backscatter behavior. We report our results on the energy distributions of the bremsstrahlung spectra; electron backscatter ratio; and spatial, temporal, and energy distributions of backscattered electrons that return to the anode.

Applying X-Ray Fluorescence With Monte Carlo Methods For Analysis Of Surface Defects On Coated Zr4, James Tyler Cahill

Theses and Dissertations

Chromium-coated Zircaloy-4 cladding represents a novel solution for creating nuclear cladding that has passive safety features. This and related developments aimed at improving nuclear safety are broadly categorized as accident-tolerant fuel and cladding. However, implementing this material presents practical challenges such as the vulnerability that cladding has to surface damage such as fretting wear. This research aims to determine if the pre-existing technique of X-ray fluorescence analysis is applicable for detecting these defects on the surface of the cladding. This was done by developing Monte Carlo N-Particle Transport models of a handheld device, the XL-5, alongside experimental measurements using this …

Modeling The Radiation Shielding Effectiveness Of Hafnium Diboride Reinforced Epoxy Composites Using Mcnp, Randolph Snook

Theses and Dissertations

This thesis uses MCNP simulations to evaluate the performance of a radiation shielding material consisting of Hafnium Diboride (HfB₂) powder in an epoxy matrix. Comparing the results to previously measured data against both gamma and neutron sources, suitability for other applications can be confirmed. The model simulated the epoxy material with different weight percentages of HfB₂ and thicknesses. It was compared to other shielding materials such as pure epoxy and 5% Borated High Density Polyethylene (HDPE-B). The HfB₂-epoxy composite was tested against a wide energy range gamma spectrum from a ¹⁵²Eu source, and neutrons …

Modeling The Reactor Time-Dependent Delayed Particle Tail With Monte Carlo N-Particle (Mcnp) Version 6.2, Eli James Boland

Masters Theses

"Energy is deposited into experiment packages due to post-shutdown decay heat created from delayed particles. Modeling these delayed particles in a reactor assists researchers in quantifying the expected energy deposition sources to an experiment package before irradiation. This paper focuses on modeling the delayed particles in a reactor in MCNP6.2 by capturing a reactor as a source, converting this source capture to a source definition, applying appropriate physics such as activation and photonuclear interactions, and finally using proper tallies to create the expected delayed particle tail of a reactor.

To capture the source distribution, the FMESH capability within MCNP was …

The Effects Of A Power Iteration-Based K-Eigenvalue Solver For Various Subcritical Parameters And Calculations, Daniel H. Timmons

Nuclear Engineering ETDs

The need to simulate subcritical benchmark parameters quickly and accurately is becoming increasing important. When using Monte Carlo methods this is traditionally done using a fixed-source calculation where a particle and its progeny are tracked until their removal from the system. This method can be slow for near critical systems. The use of a k-eigenvalue solver could reduce the computational footprint and reduce the need to post process data.

This is done for four parameters from the ICSBEP benchmark values: R_1 , R_2 , M_L, and M_eff. These parameters are calculated in two new distinct ways. First, is directly …

Aerial Radiation Detection Identification And Measurement System Detector Material Comparison Study, Benjamin C. Troxell, Kacey D. Mcgee, Christina L. Dugan

Faculty Publications

The 20th Chemical Biological Radiological Nuclear and Explosives Command (CBRNE) currently utilizes an airborne sodium iodide gamma and beta detection system to map radiation fields over large areas of interest. The 20th CBRNE explored emergent detector technologies utilizing two detection materials; thallium-activated cesium iodide and high purity germanium (HPGe). These detectors were simulated at various altitudes and compared to background measurements. The sodium iodide detector failed to provide isotopic discrimination at distance. The thallium-activated cesium iodide CsI(Tl) detector provided sufficient absolute efficiency and energy resolution to identify isotopics at distance. The HPGe detector provided the best energy resolution. However, current …

Verficiation And Validation Of Unstructured Meshing Capabilities In Mcnp6, Micah C. Jeroutek

Theses and Dissertations

Discrepancies were found between the legacy CSGs and the new UMs with the latter's introduction in MCNP6. As such, V&V of UMs in MCNP must be conducted to assist in their further development and increase user confidence in them. Currently, the MCNP test suite is lacking in UM benchmarks, and of the few that are present, their range in problem type is limited. To address these issues, two preliminary test cases covering new applications of UMs were developed for future inclusion in the test suite. The first was for the Athena-I experiment which involved the study of electronic devices' responses …

Analysis Of Non-Conventional Radiological Terrorism, James N. Padgett

Theses and Dissertations

Nuclear terrorism has been a risk since the dawn of the first atomic bomb. Though state sponsored nuclear weapons development is of concern for countries, non-state sponsored terrorism with radiological material can be of even greater concern. This stems from the fact that the material is under less stringent or no safeguards and can readily change hands between different terrorist groups or innocent civilians may accidently come into contact with the material.

Within this paper an analysis of previous accidents using orphan radiological sources, malicious use of orphan radiological sources, and how these sources could be used by terrorists is …

Monte Carlo Perturbation Analysis Of Fuel Temperature Variations In The Mcnp Model Of The Annular Core Research Reactor, Melissa Andrea Moreno

Nuclear Engineering ETDs

The Annular Core Research Reactor (ACRR) Monte Carlo N-Particle (MCNP) model is used for a variety of computational calculations ranging from reactor kinetics metrics to safety analyses. To understand the dominant source of uncertainty within the model, perturbations in temperature were applied to individual ACRR MCNP fuel rods. Assigning random temperatures, selected uniformly, from the operational temperature ranges of the fuel enables a study of uncertainty effects based on temperature variations. Stochastic mixing was used to blend the cross-sections of the desired temperatures using the MCNP continuous and Thermal Neutron Scattering Treatment (S(α,β)) libraries in ENDF/B-VII.1. The uncertainty quantification process …

Fast Neutron Assay Of Elemental Content Of Bulk Samples, Calder Emerson Lane

UNLV Theses, Dissertations, Professional Papers, and Capstones

The purpose of this research was to analyze the capabilities of fast neutrons in the detection and analysis of various isotopes in bulk samples. The deuterium-tritium (DT) fusion reaction generates highly penetrating, high-energy (14.1-MeV) neutrons which induce nuclear reactions in irradiated targets. Neutrons and gamma rays are generated in these reactions. Emitted gamma rays are characteristic of the emitter; the gamma spectrum enables stoichiometric identification of the assayed samples. Neutron backscattering can also be used for identification of the elemental composition of the sample.

This work had three objectives. The first objective was to develop a computational technique to model …

Computational Modeling Of Radiation Damage In A Multi-Phase Ceramic Waste Form Using Moose, Zeyu Chen

Theses and Dissertations

Ceramic waste forms have been proposed to replace the traditional glassy waste forms for long term stabilization of radionuclides. These waste forms are constantly exposed to self-irradiation emitted from the constituent radionuclides causing their material properties to change accordingly. It has been known that the radiation damage in waste forms is dominated by alpha particles emitted from transuranic (TRU) radionuclides. Since alpha particles usually have a range of 10~20 μm in such waste forms, some fraction of any non-transuranic containing phases (for a multiphase waste form) will be undamaged (or less damaged) if containing large enough grain sizes. Modeling and …

Characterization Of Uranium Foil Irradiations At The Wsu Triga Reactor Using A New Reactor Model In Scale, Kimberly A. Hinrichs

Nuclear Engineering ETDs

A new reactor model of the Washington State University TRIGA was developed in the SCALE neutron transport code, and its fidelity was verified by comparison to MCNP and available data for several reactor parameters. The model was used to characterize irradiations designed to produce the short-lived actinides ²³⁷U and ²³⁹U, two key isotopes for nuclear forensics. These short-lived actinides, their decay daughters ²³⁷Np and ²³⁹Pu, and total fissions (via ⁹⁹Mo) were measured in irradiated foils at Los Alamos National Laboratory and other labs with good agreement among parent/daughter pairs and among labs. The laboratory-measured isotope …

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 …

Analysis Of Energy Economy In Muon Catalyzed Fusion Considering External X-Ray Reactivation, Nishant Raghav Pillai

Masters Theses

"An analysis of the energy economy of a theoretical muon-catalyzed nuclear fusion system has been made by invoking the use of point kinetic equations, Monte Carlo radiation transport simulations, and from a review of existing literature on muon-catalyzed fusion. An external X-ray reactivation source is proposed as a novel way to increase the number of fusions per muon and thereby overcome the so-called alpha sticking problem that has long been considered the primary impediment to breakeven muon-catalyzed fusion power. Free electron lasers, synchrotrons and Wakefield accelerators are discussed as possible bright X-ray photon sources. The addition of an intense external …

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 …

Innovative Fuel Design To Improve Proliferation Management, Taylor C. Britt

Theses and Dissertations

This research uses an existing innovative fuel design (IFD) that has intrinsic safety features and enhanced economics over the current uranium dioxide (UO₂) light water fuel design and evaluates promising methods to improve the waste management and proliferation resistance of the IFD by doping the fresh fuel with select actinides.The most robust approach for proliferation resistance is to denature these materials by adding a uranium or plutonium isotope that hampers the usability of the materials in weapons. The proposed modifications to the IFD use this approach through elevated fractions of ²³⁸Pu. ²³⁸Pu generates large quantities of …

Time And Energy Characterization Of A Neutron Time Of Flight Detector Using A Novel Coincidence Method For Constraining Neutron Yield, Ion Temperature And Liner Density Measurements From Maglif Experiments, Jedediah Styron

Nuclear Engineering ETDs

The focus of this work is the characterization of a typical neutron time-of-flight (NTOF) detector that is fielded on inertial confinement fusion (ICF) experiments conducted at the Z-experimental facility with emphasis on the Magnetized Liner Fusion (MagLIF) concept. An NTOF detector consisting of EJ-228 scintillator and two independent photomultiplier tubes (PMTs), a Hamamatsu-mod 5 and Photek-PMT240, has been characterized in terms of the absolute time and energy response. The characterization was done by measuring single, neutron-induced events in the scintillator by measuring the alpha particle and neutron produced from the D-T reaction in kinematic coincidence. The results of these experiments …

A Study Of The Potential Applications Of Am241, And Determining The Feasibility Of Using Gamma Spectroscopy For Future Physical Validation, Eric A. Feissle

Masters Theses

“Am241 is typically produced via Pu241 decay in a uranium fueled reactor. Presence of Am241 can be used as the age estimation tool for spent fuel, which is a focus of this thesis along with the interest of the measurement and the ratio of production rates of Am241’s activation products; Americium-242 and its first excited state of Americium-242m. MCNP models of the core and BEGe 3825 detector were built. These models were compared with established and physical measurements of gamma/x-ray standards that were available at the reactor. Thermal fluxes at 200 kW for potential foils centered in the source holder …

The Viability Of Advantg Deterministic Method For Synthetic Radiography Generation, Andrew Albert Bingham

Masters Theses

"Time sensitive and high resolution image simulations are needed for synthetic radiography generation. The standard stochastic approach requires lengthy run times with poor statistics at higher resolutions. The investigation of the viability of a deterministic approach to synthetic radiography image generation was explored. The aim was to analyze a computational time decrease over the stochastic method. ADVANTG was compared to MCNP in multiple scenarios including a Benchtop CT prototype, to simulate high resolution radiography images. By using ADVANTG deterministic code to simulate radiography images the computational time was found to decrease over 10 times compared to the MCNP stochastic approach"--Abstract, …

Doppler Temperature Coefficient Calculations Using Adjoint-Weighted Tallies And Continuous-Energy Cross Sections In Mcnp6, Matthew A. Gonzales

Nuclear Engineering ETDs

The calculation of the thermal neutron Doppler temperature reactivity feedback co- efficient, a key parameter in the design and safe operation of advanced reactors, using first order perturbation theory in continuous energy Monte Carlo codes is challenging as the continuous energy adjoint flux is not readily available. Traditional approaches of obtaining the adjoint flux attempt to invert the random walk process as well as require data corresponding to all temperatures and their respective tem- perature derivatives within the system in order to accurately calculate the Doppler temperature feedback.

A new method has been developed using adjoint-weighted tallies and On-The-Fly (OTF) …

Mcnp6 Computational-Based Sensitivity Propagation Analysis Of Continuous Neutron Cross-Sections Using The Godiva (Hmf-001) And The Godiver (Hmf-004) Benchmark Criticality Study Cases, Lawrence James Lakeotes

UNLV Theses, Dissertations, Professional Papers, and Capstones

There has been a reduction in funding for theoretical and applied research for improving the nation's database of continuous neutron cross-sections at BNL-NNDC. From 1940 through the late 1980s, research and applied development produced volumes of reliable neutron continuous cross-sections for many isotopes. Currently, the cross-section work has been mainly computational. The focus of this research is mainly centered on the requirements for improving thermal cross-sections to support reactor operations and fuel storage. The research efforts will also helpfully aid in the fast fission spectrum in order to support fast reactor designs for improving safety analysis and feedback coefficients.

This …

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 …

Attenuation Properties Of Cement Composites: Experimental Measurements And Monte Carlo Calculations, Raul Florez

Masters Theses

"Developing new cement based materials with excellent mechanical and attenuation properties is critically important for both medical and nuclear power industries. Concrete continues to be the primary choice material for the shielding of gamma and neutron radiation in facilities such as nuclear reactors, nuclear waste repositories, spent nuclear fuel pools, heavy particle radiotherapy rooms, particles accelerators, among others. The purpose of this research was to manufacture cement pastes modified with magnetite and samarium oxide and evaluate the feasibility of utilizing them for shielding of gamma and neutron radiation. Two different experiments were conducted to accomplish these goals. In the first …

Impact Of Configuration Variations On Small Modular Reactor Core Performance, William Kirby Compton

Masters Theses

"One of the most promising new reactor designs is the Small Modular Reactor (SMR). These reactors, which operate under 300 MWe, will help bring cheap and safe nuclear energy to remote and centralized locations alike. Their ease of construction, advanced passive safety features, and cost effectiveness make these reactors an intriguing option for the near future.

In the work presented here, a neutronics analysis of the Westinghouse SMR was performed. Westinghouse's SMR design is a scaled down version of their AP1000 plant and will produce about 225 MWe of power. Though the parameters of the reactor core will be modeled …

Verification Of A Monte Carlo Model Of The Missouri S&T Reactor, Brad Paul Richardson

Masters Theses

The purpose of this research is to ensure that an MCNP model of the Missouri S&T reactor produces accurate results so that it may be used to predict the effects of some desired upgrades to the reactor. The desired upgrades are an increase in licensed power from 200 kW to 400kW, and the installation of a secondary cooling system to prevent heating of the pool. This was performed by comparing simulations performed using the model with experiments performed using the reactor. The experiments performed were, the approach to criticality method of predicting the critical control rod height, measurement of the …