VTechWorks Repository :: Browsing by Author "Pierson, Mark Alan"

Browsing by Author "Pierson, Mark Alan"

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Analysis and Improvement of the bRAPID Algorithm and its Implementation
Bartel, Jacob Benjamin (Virginia Tech, 2019-07-18)
This thesis presents a detailed analysis of the bRAPID (burnup for RAPID – Real Time Analysis for Particle transport and In-situ Detection) code system, and the implementation and validation of two new algorithms for improved burnup simulation. bRAPID is a fuel burnup algorithm capable of performing full core 3D assembly-wise burnup calculations in real time, through its use of the RAPID Fission Matrix methodology. A study into the eﬀect of time step resolution on isotopic composition in Monte Carlo burnup calculations is presented to provide recommendations for time step scheme development in bRAPID. Two novel algorithms are implemented into bRAPID, which address: i) the generation of time-dependent correction factors for the ﬁssion density distribution in boundary nuclear fuel assemblies within a reactor core; ii) the calculation of pin-wise burnup distributions and isotopic concentrations. Time step resolution analysis shows that a variable time step scheme, developed to accurately characterize important isotope evolution, can be used to optimize burnup calculations and minimize computation time. The two new algorithms have been benchmarked against the Monte Carlo code system Serpent. Results indicate that the time-dependent boundary correction algorithm improves ﬁssion density distribution calculations by including a more detailed representation of boundary physics. The pin-wise burnup algorithm expands bRAPID capabilities to provide material composition data at the pin level, with accuracy comparable to the reference calculation. In addition, wall-clock time analyses show that burnup calculations performed using bRAPID with these novel algorithms require a fraction of the time of Serpent.
Autonomous Source Localization
Peterson, John Ryan (Virginia Tech, 2020-05-01)
This work discusses the algorithms and implementation of a multi-robot system for locating radioactive sources. The estimation algorithm presented in this work is able to fuse measurements collected by γ-ray spectrometers carried by an unmanned aerial and unmanned ground vehicle into a single consistent estimate of the probability distribution over the position of a point source in an environment. By constructing a set of hypotheses on the position of the point source, this method converts a non-linear problem into many independent linear ones. Since the underlying model is probabilistic, candidate paths may be evaluated by their expected reduction in uncertainty, allowing the algorithm to select good paths for vehicles to take. An initial hardware test conducted at Savannah River National Laboratory served as a proof of concept and demonstrated that the algorithm successfully locates a radioactive source in the environment, and moves the vehicle to that location. This approach also demonstrated the capability to utilize radiation data collected from an unmanned aerial vehicle to aid the ground vehicle’s exploration. Subsequent numerical experiments characterized the performance of several reward functions and different exploration algorithms in scenarios covering a range of source strengths and region sizes. These experiments demonstrated the improved performance of planning-based algorithms over the myopic method initially tested in the hardware experiments.
Benchmarking of the RAPID Eigenvalue Algorithm using the ICSBEP Handbook
Butler, James Michael (Virginia Tech, 2019-09-17)
The purpose of this thesis is to examine the accuracy of the RAPID (Real-Time Analysis for Particle Transport and In-situ Detection) eigenvalue algorithm based on a few problems from the ICSBEP (International Criticality Safety Benchmark Evaluation Project) Handbook. RAPID is developed based on the MRT (Multi-Stage Response-Function Transport) methodology and it uses the fission matrix (FM) method for performing eigenvalue calculations. RAPID has already been benchmarked based on several real-world problems including spent fuel pools and casks, and reactor cores. This thesis examines the accuracy of the RAPID eigenvalue algorithm for modeling the physics of problems with unique geometric configurations. Four problems were selected from the ICSBEP Handbook; these problems differ by their unique configurations which can effectively examine the capability of the RAPID code system. For each problem, a reference Serpent Monte Carlo calculation has been performed. Using the same Serpent model in the pRAPID (pre- and post-processing for RAPID) utility code, a series of fixed-source Serpent calculations are performed to determine spatially-dependent FM coefficients. RAPID calculations are performed using these FM coefficients to obtain the axially-dependent, pin-wise fission density distribution and system eigenvalue for each problem. It is demonstrated that the eigenvalues calculated by RAPID and Serpent agree with the experimental data within the given experimental uncertainty. Further, the detailed 3-D pin-wise fission density distribution obtained by RAPID agrees with the reference prediction by Serpent which itself has converged to less than 1% weighted uncertainty. While achieving accurate results, RAPID calculations are significantly faster than the reference Serpent calculations, with a calculation time speed-up of between 4x and 34x demonstrated in this thesis. In addition to examining the accuracy of the RAPID algorithm, this thesis provides useful information on the use of the FM method for simulation of nuclear systems.
Carbon Nanotube Based Dosimetry of Neutron and Gamma Radiation
Nelson, Anthony J. (Virginia Tech, 2016-04-29)
As the world's nuclear reactors approach the end of their originally planned lifetimes and seek license extensions, which would allow them to operate for another 20 years, accurate information regarding neutron radiation exposure is more important than ever. Structural components such as the reactor pressure vessel (RPV) become embrittled by neutron irradiation, reducing their capability to resist crack growth and increasing the risk of catastrophic failure. The current dosimetry approaches used in these high flux environments do not provide real-time information. Instead, radiation dose is calculated using computer simulations, which are checked against dose readings that are only available during refueling once every 1.5-2 years. These dose readings are also very expensive, requiring highly trained technicians to handle radioactive material and operate specialized characterization equipment. This dissertation describes the development of a novel neutron radiation dosimeter based on carbon nanotubes (CNTs) that not only provides accurate real-time dosimetry, but also does so at very low cost, without the need for complex instrumentation, highly trained operators, or handling of radioactive material. Furthermore, since this device is based on radiation damage rather than radioactivation, its readings are time-independent, which is beneficial for nuclear forensics. In addition to development of a novel dosimeter, this work also provides insight into the particularly under-investigated topic of the effects of neutron irradiation of carbon nanotubes. This work details the fabrication and characterization of carbon nanotube based neutron and gamma radiation dosimeters. They consist of a random network of CNTs, sealed under a layer of silicon dioxide, spanning the gap between two electrodes to form a conductive path. They were fabricated using conventional wafer processing techniques, making them intrinsically scalable and ready for mass production. Electrical properties were measured before and after irradiation at several doses, demonstrating a consistent repeatable trend that can be effectively used to measure dose. Changes to the microstructure were investigated using Raman spectroscopy, which confirmed that the changes to electrical properties are due to increasing defect concentration. The results outlined in this dissertation will have significant impacts on both the commercial nuclear industry and on the nanomaterials scientific community. The dosimeter design has been refined to the point where it is nearly ready to be deployed commercially. This device will significantly improve accuracy of RPV lifetime assessment while at the same time reducing costs. The insights into the behavior of CNTs in neutron and gamma radiation environments is of great interest to scientists and engineers studying these nanomaterials.
CFD Analysis of Aspirator Region in a B&W Enhanced Once-Through Steam Generator
Spontarelli, Adam Michael (Virginia Tech, 2013-06-07)
This analysis calculates the velocity profile and recirculation ratio in the aspirator region of an enhanced once-through steam generator of the Babcock & Wilcox design. This information is important to the development of accurate RELAP5 models, steam generator level calculations, steam generator downcomer models, and flow induced vibration analyses. The OpenFOAM CFD software package was used to develop the three-dimensional model of the EOTSG aspirator region, perform the calculations, and post-process the results. Through a series of cases, each improving upon the modeling accuracy of the previous, insight is gained into the importance of various modeling considerations, as well as the thermal-hydraulic behavior in the steam generator downcomer. Modeling the tube support plates and tube nest is important for the accurate prediction of flow rates above and below the aspirator port, but has little affect on the aspirator region itself. Modeling the MFW nozzle has minimal influence on the incoming steam velocity, but does create a slight azimuthal asymmetry and alter the flow pattern in the downcomer, creating recirculation patterns important to inter-phase heat transfer. Through the development of a two-phase solution that couples the aspirated steam and liquid feedwater, it was found that the ratio of droplet surface area to volume plays the most important role in determining the rate of aspiration. Calculations of the velocity profile and recirculation ratio are compared against those of historical calculations, demonstrating the possibility that these parameters were previously underpredicted. Such a conclusion can only be confidently made once experimental data is made available to validate the results of this analysis.
A Computational Study of A Lithium Deuteride Fueled Electrothermal Plasma Mass Accelerator
Gebhart, Gerald Edward III (Virginia Tech, 2013-06-13)
Future magnetic fusion reactors such as tokamaks will need innovative, fast, deep-fueling systems to inject frozen deuterium-tritium pellets at high speeds and high repetition rates into the hot plasma core. There have been several studies and concepts for pellet injectors generated, and different devices have been proposed. In addition to fueling, recent studies show that it may be possible to disrupt edge localized mode (ELM) formation by injecting pellets or gas into the fusion plasma. The system studied is capable of doing either at a variety of plasma and pellet velocities, volumes, and repetition rates that can be controlled through the formation conditions of the plasma. In magnetic or inertial fusion reactors, hydrogen, its isotopes, and lithium are used as fusion fueling materials. Lithium is considered a fusion fuel and not an impurity in fusion reactors as it can be used to produce fusion energy and breed fusion products. Lithium hydride and lithium deuteride may serve as good ablating sleeves for plasma formation in an ablation-dominated electrothermal plasma source to propel fusion pellets. Previous studies have shown that pellet exit velocities, greater 3 km/s, are possible using low-z propellant materials. In this work, a comprehensive study of solid lithium hydride and deuteride as a pellet propellant is conducted using the ETFLOW code, and relationships between propellants, source and barrel geometry, pellet volume and aspect ratio, and pellet velocity are determined for pellets ranging in volume from 1 to 100 mm3.
Design and Characterization of a Coaxial Plasma Railgun for Jet Collision Experiments
Coleman, Mathew Riley (Virginia Tech, 2021-03-17)
Plasma railguns are electromagnetic accelerators used to produce controlled high velocity plasma jets. This thesis discusses the design and characterization of a small coaxial plasma railgun intended to accelerate argon-helium plasma jets. The railgun will be used for the study of plasma shocks in jet collisions. The railgun is mounted on a KF-40 vacuum port and operated using a 90 kA, 11 kV LC pulse forming network. Existing knowledge of coaxial railgun plasma instabilities and material interactions at vacuum and plasma interfaces are applied to the design. The design of individual gun components is detailed. Jet velocity and density are characterized by analyzing diagnostic data collected from a Rogowski coil, interferometer, and photodiode. Peak line-integrated electron number densities of approximately 8 × 10¹⁵ cm^-2 and jet velocities of tens of km/s are inferred from the data recorded from ten experimental pulses.
Design of Optical Measurements for Electrothermal Plasma Discharges
Hamer, Matthew David (Virginia Tech, 2014-06-23)
Ablation controlled electrothermal (ET) plasma discharge devices consist of a small diameter capillary through which a large amount of energy is discharged. The high energy in the discharge ablates an inner sleeve material, ionizes the material, and a high energy-density plasma jet accelerates out the open end. ET devices can find applications in internal combustion engines, Tokamak fusion fueling and stabilization, hypervelocity launchers, and propulsion. The ballistic properties of an ET device are highly dependent on the propellant and ablated material. A useful noninvasive technique to characterize a propellant in these types of devices is spectroscopy. The purpose of this study is to design and conduct experiments on the ET facility called PIPE to verify results and assumptions in the ETFLOW simulation code as well as resolve data collection issues such as equipment triggering as spectrometer saturation. Experiments are carried out using an Ocean Optics LIBS2500plus high resolution spectrometer and a Photron FASTCAM SA4 high speed camera. Electron plasma temperatures are estimated using copper peaks in the UV region with the relative line intensity method, and electron plasma density is estimated by measuring the full width at half maximum (FWHM) of the stark broadened H--β line at 486 nm. Electron temperatures between 0.19 eV and 0.49 eV, and electron densities between 4.68*1022 m-3 and 5.75*10²² m⁻³ were measured in the expanding plasma jet about an inch outside the source with values as expected for this region. Velocity measurements of PIPE match well with simulations at around 5333 m/s. This study concluded that the assumption that the propellant Lexan is completely dissociated is a valid assumption, and that the ETFLOW results for electron temperature, density, and bulk plasma velocity match experimental values.
Design of Optical Measurements for Plasma Actuators for the Validation of Quiescent and Flow Control Simulations
Lam, Derrick Chuk-Wung (Virginia Tech, 2016-01-27)
The concept of plasma flow control is a relatively new idea based on using atmospheric plasma placed near the edge of an air foil to reduce boundary layer losses. As with any new concept, it is important to be able to quantify theoretical assumptions with known experimental results for validation. Currently there are a variety of experiments being done to better understand plasma flow control, but one particular experiment is through the use of multi-physics modeling of dielectric barrier discharge actuators. The research in this thesis uses optical measurement techniques to validate computational models of flow control actuators being done concurrently at Virginia Tech. The primary focus of this work is to design, build and test plasma actuators in order to determine the plasma characteristics relating to electron temperatures and densities. Using optical measurement techniques such as plasma spectroscopy, measured electron temperatures and densities to compare with theoretical calculations of plasma flow control under a variety of flow conditions. This thesis covers a background of plasma physics, optical measurement techniques, and the designing of the plasma actuator setups used in measuring atmospheric plasmas.
Desing of the high Pressure HIgh temperature annuLUS flow (PHILUS) Facility
Karabacak, Ali Haydar (Virginia Tech, 2022-06-17)
Critical heat flux (CHF) and post-CHF are two critical phenomena in light water-cooled nuclear power plants regarding safety. Even though the general trends of CHF and post- CHF are known, the exact mechanisms are still unknown. To better understand CHF and post-CHF, experimental flow boiling facilities are constructed around the world. However, these facilities are limited in their experimental conditions and spatial resolution necessary to advance our understanding of two-phase heat transfer. Previous rod surface measurements were collected with thermocouples to measure CHF location and temperature excursion, yet thermocouples provide limited spatial resolution, which leads to significant uncertainties in the CHF prediction. On the other hand, optical fiber temperature sensors can measure the temperature and the CHF propagation with high spatial resolution. Also, the capability of the optical fiber at high temperatures has been proven in previous studies. The current study aims to apply optical fiber at high-pressure and high mass fluxes. The high-Pressure HIgh-temperature annuLUS flow (PHILUS) facility was designed to provide desired working conditions in the test section that uses optical fiber temperature sensors. The PHILUS test section has a length of 1320 mm, with 1000 mm of heated length. The working conditions of the PHILUS are up to 18 MPa, temperatures up to 357◦C, and coolant mass flux from 500 to 3700 kg/m2s. The main components of the loop are a steam separator, two heat exchangers (a condenser and a cooler), a bladder-type accumulator, two bypass lines, and a high-pressure pump. Coolant-Boiling in Rod Arrays-Two Fluids (COBRA-TF) code was used to design the CHF and post-CHF experiments to be performed at the PHILUS facility.
Development of a Minichannel Compact Primary Heat Exchanger for a Molten Salt Reactor
Lippy, Matthew Stephen (Virginia Tech, 2011-04-28)
The first Molten Salt Reactor (MSR) was designed and tested at Oak Ridge National Laboratory (ORNL) in the 1960's, but recent technological advancements now allow for new components, such as heat exchangers, to be created for the next generation of MSR's and molten salt-cooled reactors. The primary (fuel salt-to-secondary salt) heat exchanger (PHX) design is shown here to make dramatic improvements over traditional shell-and-tube heat exchangers when changed to a compact heat exchanger design. While this paper focuses on the application of compact heat exchangers on a Molten Salt Reactor, many of the analyses and results are similarly applicable to other fluid-to-fluid heat xchangers. The heat exchanger design in this study seeks to find a middle-ground between shell- and-tube designs and new ultra-efficient, ultra-compact designs. Complex channel geometries and microscale dimensions in modern compact heat exchangers do not allow routine maintenance to be performed by standard procedures, so extended surfaces will be omitted and hydraulic diameters will be kept in the minichannel regime (minimum channel dimension between 200 μm and 3 mm) to allow for high-frequency eddy current inspection methods to be developed. High aspect ratio rectangular channel cross-sections are used. Various plant layouts of smaller heat exchanger banks in a "modular" design are introduced. FLUENT was used within ANSYS Workbench to find optimized heat transfer and hydrodynamic performance. With similar boundary conditions to ORNL's Molten Salt Breeder Reactor's shell-and-tube design, the compact heat exchanger interest in this thesis will lessen volume requirements, lower fuel salt volume, and decrease material usage.
Development of a Novel Fuel Burnup Methodology and Algorithm in RAPID and its Benchmarking and Automation
Roskoff, Nathan (Virginia Tech, 2018-08-02)
Fuel burnup calculations provide material concentrations and intrinsic neutron and gamma source strengths as a function of irradiation and cooling time. Detailed, full-core 3D burnup calculations are critical for nuclear fuel management studies, including core design and spent fuel storage safety and safeguards analysis. For core design, specifically during refueling, full- core pin-wise, axially-dependent burnup distributions are necessary to determine assembly positioning to efficiently utilize fuel resources. In spent fuel storage criticality safety analysis, detailed burnup distributions enable best-estimate analysis which allows for more effective utilization of storage space. Additionally, detailed knowledge of neutron and gamma source distributions provide the ability to ensure nuclear material safeguards. The need for accurate and efficient burnup calculations has become more urgent for the simulation of advanced reactors and monitoring and safeguards of spent fuel pools. To this end, the Virginia Tech Transport Theory Group (VT3G) has been working on advanced computational tools for accurate modeling and simulation of nuclear systems in real-time. These tools are based on the Multi-stage Response-function Transport (MRT) methodology. For monitoring and safety evaluation of spent fuel pools and casks, the RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system has been developed. This dissertation presents a novel methodology and algorithm for performing 3D fuel bur- nup calculations, referred to as bRAPID- Burnup with RAPID . bRAPID utilizes the existing RAPID code system for accurate calculation of 3D fission source distributions as the trans- port calculation tool to drive the 3D burnup calculation. bRAPID is capable of accurately and efficiently calculating assembly-wise axially-dependent fission source and burnup dis- tributions, and irradiated-fuel properties including material compositions, neutron source, gamma source, spontaneous fission source, and activities. bRAPID performs 3D burnup calculations in a fraction of the time required by state-of-the-art methodologies because it utilizes a pre-calculated database of response functions. The bRAPID database pre-calculation procedure, and its automation, is presented. The ex- isting RAPID code is then benchmarked against the MCNP and Serpent Monte Carlo codes for a spent fuel pool and the U.S. Naval Academy Subcritical Reactor facility. RAPID is shown to accurately calculate eigenvalue, subcritical multiplication, and 3D fission source dis- tributions. Finally, bRAPID is compared to traditional, state-of-the art Serpent Monte Carlo burnup calculations and its performance will be evaluated. It is important to note that the automated pre-calculation proceedure is required for evaluating the performance of bRAPID. Additionally, benchmarking of the RAPID code is necessary to understand RAPID's ability to solve problems with variable burnups distributions and to asses its accuracy.
Development of Advanced Image Processing Algorithms for Bubbly Flow Measurement
Fu, Yucheng (Virginia Tech, 2018-10-16)
An accurate measurement of bubbly flow has a significant value for understanding the bubble behavior, heat and energy transfer pattern in different engineering systems. It also helps to advance the theoretical model development in two-phase flow study. Due to the interaction between the gas and liquid phase, the flow patterns are complicated in recorded image data. The segmentation and reconstruction of overlapping bubbles in these images is a challenging task. This dissertation provides a complete set of image processing algorithms for bubbly flow measurement. The developed algorithm can deal with bubble overlapping issues and reconstruct bubble outline in 2D high speed images under a wide void fraction range. Key bubbly flow parameters such as void fraction, interfacial area concentration, bubble number density and velocity can be computed automatically after bubble segmentation. The time-averaged bubbly flow distributions are generated based on the extracted parameters for flow characteristic study. A 3D imaging system is developed for 3D bubble reconstruction. The proposed 3D reconstruction algorithm can restore the bubble shape in a time sequence for accurate flow visualization with minimum assumptions. The 3D reconstruction algorithm shows an error of less than 2% in volume measurement compared to the syringe reading. Finally, a new image synthesis framework called Bubble Generative Adversarial Networks (BubGAN) is proposed by combining the conventional image processing algorithm and deep learning technique. This framework aims to provide a generic benchmark tool for assessing the performance of the existed image processing algorithms with significant quality improvement in synthetic bubbly flow image generation.
Development of High-Speed Camera Techniques for Droplet Measurement in Annular Flows
Cohn, Ayden Seth (Virginia Tech, 2024-06-03)
This research addresses the critical need for precise two-phase flow data in the development of computer simulation models, with a specific focus on the annular flow regime's droplet behavior. The study aims to contribute to the evaluation of safety and efficiency in nuclear reactors that handle fluids transitioning between liquid and gas states for thermal energy transport. Central to the investigation is the collection and analysis of droplet size and velocity distribution data, particularly to help with developing models for the water-cooled nuclear power plants. The experimental setup employs advanced tools, including a high-speed camera, lens, teleconverter, and a selected light source, to capture high-resolution images of droplets. Calibration procedures, incorporating depth of field testing, are implemented to ensure accurate droplet size measurements. A critical component of the research is the introduction of a droplet identification program, developed using Matlab, which facilitates efficient processing of experimental data. Preliminary results from the Virginia Tech test facility demonstrate the system's capability to eliminate out-of-focus droplets and obtain precise droplet data in a reasonable amount of time. Experimental results from the Rensselaer Polytechnic Institute test facility provide droplet size and velocity distributions for a variety of annular flow conditions. This facility has a concurrent two-flow system that pumps air and water at different rates through a 9.525 mm inner diameter tube. The conditions tested include gas superficial velocities ranging from 22 to 40 m/s and liquid superficial velocities ranging from 0.09 to 0.44 m/s. The measured flow has a temperature of 21°C and a pressure of 1 atm.
Development of Metallic Fuel Additives and Alloys for Sodium-cooled Fast Reactors
Zhuo, Weiqian (Virginia Tech, 2022-07-11)
The major goal of the work is to develop effective additives for U-10Zr (wt.%) metallic fuel to mitigate the fuel-cladding chemical interactions (FCCIs) due to fission product lanthanides and to optimize the fuel phase mainly by lowering the gamma-onset temperature. The additives Sb, Mo, Nb, and Ti have been investigated. Metallic fuels with one or two of the additives and with or without lanthanide fission products were fabricated. In this study, Ce was selected as the representative lanthanide fission product. A series of tests and characterizations were carried out on the additive-bearing fuels, including annealing, diffusion coupling, scanning electron microscopy (SEM), X-ray powder diffraction (XRD), and differential scanning calorimetry (DSC). Sb was investigated to mitigate FCCIs because available studies show its potential as a lanthanide immobilizer. This work extends the knowledge of Sb in U-10Zr, including its effect in the Zr-free region. Sb forms precipitates with fuel constituents, either U or Zr. However, it combines with the lanthanide fission product Ce when Ce is present. Those Sb-precipitates are found to be stable upon annealing, and are compatible with the cladding. The additive does not change the phase transition of U-10Zr. Mo, Nb, and Ti have been investigated for phase optimization based on the known characteristics shown in the binary phase diagrams. The quaternary alloys, i.e., two Mo-bearing alloys and two Nb-bearing alloys, were investigated. Compared to U-10Zr, a few weight percentages of Zr are replaced by those additives in the quarternary alloys. The solid-state phase transitions were determined (alpha and U2Ti transfer into gamma). The transition temperature varies depending on the compositions. The Mo-bearing alloys have lower -onset temperatures than the Nb-bearing alloys. All of them have lower gamma-onset temperatures than that of U-10Zr. Since low gamma-onset temperature is favorable, the results indicate that the fuel phase can be optimized by the replacement of a few weight percentages of Zr into those additives. All the experiments were out-of-pile tests. Therefore, in-pile experiments will be necessary to fully evaluate the performance of the additives in the future.
Development of the Adaptive Collision Source Method for Discrete Ordinates Radiation Transport
Walters, William Jonathan (Virginia Tech, 2015-05-08)
A novel collision source method has been developed to solve the Linear Boltzmann Equation (LBE) more efficiently by adaptation of the angular quadrature order. The angular adaptation method is unique in that the flux from each scattering source iteration is obtained, with potentially a different quadrature order used for each. Traditionally, the flux from every iteration is combined, with the same quadrature applied to the combined flux. Since the scattering process tends to distribute the radiation more evenly over angles (i.e., make it more isotropic), the quadrature requirements generally decrease with each iteration. This method allows for an optimal use of processing power, by using a high order quadrature for the first few iterations that need it, before shifting to lower order quadratures for the remaining iterations. This is essentially an extension of the first collision source method, and is referred to as the adaptive collision source (ACS) method. The ACS methodology has been implemented in the 3-D, parallel, multigroup discrete ordinates code TITAN. This code was tested on a variety of test problems including fixed-source and eigenvalue problems. The ACS implementation in TITAN has shown a reduction in computation time by a factor of 1.5-4 on the fixed-source test problems, for the same desired level of accuracy, as compared to the standard TITAN code.
Discrete-Time Bayesian Networks Applied to Reliability of Flexible Coping Strategies of Nuclear Power Plants
Sahin, Elvan (Virginia Tech, 2021-06-11)
The Fukushima Daiichi accident prompted the nuclear community to find a new solution to reduce the risky situations in nuclear power plants (NPPs) due to beyond-design-basis external events (BDBEEs). An implementation guide for diverse and flexible coping strategies (FLEX) has been presented by Nuclear Energy Institute (NEI) to manage the challenge of BDBEEs and to enhance reactor safety against extended station blackout (SBO). To assess the effectiveness of FLEX strategies, probabilistic risk assessment (PRA) methods can be used to calculate the reliability of such systems. Due to the uniqueness of FLEX systems, these systems can potentially carry dependencies among components not commonly modeled in NPPs. Therefore, a suitable method is needed to analyze the reliability of FLEX systems in nuclear reactors. This thesis investigates the effectiveness and applicability of Bayesian networks (BNs) and Discrete-Time Bayesian Networks (DTBNs) in the reliability analysis of FLEX equipment that is utilized to reduce the risk in nuclear power plants. To this end, the thesis compares BNs with two other reliability assessment methods: Fault Tree (FT) and Markov chain (MC). Also, it is shown that these two methods can be transformed into BN to perform the reliability analysis of FLEX systems. The comparison of the three reliability methods is shown and discussed in three different applications. The results show that BNs are not only a powerful method in modeling FLEX strategies, but it is also an effective technique to perform reliability analysis of FLEX equipment in nuclear power plants.
The Effects of Neutron and Gamma Radiation on Graphene
Kryworuk, Christopher Nicholas (Virginia Tech, 2013-06-03)
Although young in its existence, graphene has already shown many potential uses in nuclear engineering. Graphene has unique electrical, mechanical and optical properties that give it unmatched potential for applications raging from sensors to composites. Before these applications can be fully developed, the response to neutron and gamma irradiation must be understood. In this study, graphene grown from chemical vapor deposition was irradiated by the High Flux Isotope Reactor at Oak Ridge National Laboratory and characterized using Raman spectroscopy. It was found that the amount of structural damage was minimal, but that the graphene was doped reversibly with H₂0₂ and irreversibly. The irreversible doping is a type of soft etching process related to the exposure to O₂ as well as ionizations and heating caused by irradiation. The reversible doping is related to the products generated through the radiolysis of the water trapped between the sample and the substrate. By removing the water through evaporation the dopants related to the radiolysis products were found to be removed as well. These results are promising as they show that graphene is resilient and sensitive to the effects of irradiation simultaneously.
Effects of Nodalization on Containment Analysis in a Loss of Coolant Accident Using GOTHIC
McNeil, Wilfred J. IV (Virginia Tech, 2013-05-21)
Existing containment models for a loss of coolant accident at many nuclear power plants were created in the 1970s using older computer technology and thermal hydraulic models which were available at that time. While conservative, these models may not present the detail necessary to identify conditions which may be used to produce additional design margin for the plant. After exploring containment and critical flow modeling, the basis for the use of GOTHIC in this analysis was established. A GOTHIC model was then created to simulate the loss of coolant accident results shown in an Updated Final Safety Analysis Report analysis for the North Anna Power Station. This model was used to examine the effects of increased nodalization in a subcompartment on the existing containment model. It is shown that adding multidimensional sub-nodes to areas of interest can provide valuable detail which was absent in the UFSAR model. Simulations are able to show the localized pressure spike around a LOCA pipe break that quickly dissipates, leaving significantly lower pressures in what was once an averaged, single, lumped-parameter node. This suggests that additional design margin may exist depending on where the pipe break is assumed to occur.
Effects of Proton Irradiation on the Mechanical and Physical Properties of Carbon Nanotube Based Composites
Nelson, Anthony J. (Virginia Tech, 2014-01-27)
In this study, the effects of proton irradiation on carbon nanotube (CNT)-epoxy composites are investigated for potential applications in radiation shielding for spacecraft. CNT-epoxy composites were prepared using multiwall and single wall CNTs and exposed to proton beams of energies ranging from 6 MeV to 12 MeV. The nanocomposites shielding capabilities against the different energetic proton beams were measured by tracking the beam's energy before and after penetrating the samples. The microstructures of the samples were characterized using scanning electron microscopy (FESEM). The effect of proton irradiation on the electrical resistivity was measured using a high resolution multimeter. Finally the influence of the irradiation on the mechanical properties, such as the elastic modulus and hardness, was probed using instrumented nanoindentation tests. The proton stopping power of the epoxy was shown to be unchanged by the addition of CNTs, which is a promising result since the hardness of the samples was shown to be increased by addition of CNTs. Unfortunately, however, the surface of the samples proved to be too rough for nanoindentation to yield more detailed results. This was due to the use of a diamond saw in cutting the samples to size. The addition of CNTs was shown to reduce the volume electrical resistivity of the neat epoxy by almost five orders of magnitude and the irradiation further reduced it by a factor of 2-16.

Browsing by Author "Pierson, Mark Alan"

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