A. James Clark School of Engineering
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The collections in this community comprise faculty research works, as well as graduate theses and dissertations.
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Item Insulating Materials for an Extreme Environment in a Supersonically Rotating Fusion Plasma(2024) Schwartz, Nick Raoul; Koeth, Timothy W; Material Science and Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md.)Fusion energy has long been sought as the “holy grail” of energy sources. One of the most critical remaining challenges in fusion is that of plasma-facing materials, even denoted by the National Academies of Science. The materials challenge is particularly acute for centrifugal mirrors, an alternative concept to the industry-standard tokamak that may offer a more efficient scheme with a faster path to development. The centrifugal mirror incorporates supersonic rotation into a conventional magnetic mirror scheme, providing three primary benefits: (1) increased confinement, (2) suppression of instabilities, and (3) plasma heating through shear flow. However, this rotation, which is driven by an axial magnetic field and a radial electric field, requires the magnetic field lines to terminate on electrically insulating surfaces to avoid “shorting” the plasma. This unique requirement presents a novel materials challenge: the insulator must not only resist irradiation and thermal damage, but also be an excellent electrical insulator and thermal conductor that can be actively cooled. To address this materials challenge, the Centrifugal Mirror Fusion Experiment (CMFX) was developed at the University of Maryland. CMFX serves as a test bed for electrically insulating materials in a fusion environment, as well as a proof-of-concept for the centrifugal mirror scheme. To guide the design of future power plants and better understand the neutronand ion flux on the insulators, a zero-dimensional (0-D) scoping tool, called MCTrans++, was developed. This software, discussed in Chapter 2, demonstrates the ability to rapidly model experimental parameter sets in CMFX and predict the scaling to larger devices, informing material selection and design. Assuming the engineering challenges have been met, the centrifugal mirror has been demonstrated as a promising scheme for electricity production via fusion energy. One of the key aspects to the operation of CMFX is the high voltage system. This system, discussed in Chapter 3, was developed in incremental stages, beginning with a 20 kV, then 50 kV pulsed power configuration, and finally culminating in a 100 kV direct current power supply to drive rotation at much higher voltages, creating an extreme environment for materials testing. This work identified hexagonal boron nitride (hBN) as a promising insulator material. Computational modeling (Chapter 4) demonstrated hBN’s superior resistance to ion-irradiation damage compared to other plasma-facing materials. Additionally, fusion neutrons are crucial for assessing both material damage and power output. Chapter 5 details the neutronics for CMFX, including 3He proportional counters, which have been installed on CMFX to measure neutron production. In parallel, Monte Carlo computational methods were used to predict neutron transport and material damage in the experiment. Ultimately, a materials test stand was installed on CMFX to expose electrically insulating materials to high energy fusion plasmas (Chapter 6). Comparative analysis of hBN and silicon carbide after exposure revealed superior performance of hBN as a plasma-facing material. Two primary erosion mechanisms were identified by surface morphology and roughness measurements: grain ejection and sputtering, both more pronounced in silicon carbide. This work advances our understanding of insulating material behavior in fusion environments and paves the way for the development of the next-generation centrifugal mirror fusion reactors. Chapter 7 discusses conclusions and proposes future work. In particular this section suggests some changes that may allow CMFX to operate at much higher voltages, unlocking higher plasma density and temperature regimes for further material testing.Item Dynamic Bayesian Network Data Updating Approaches for Enabling Causal Prognostics and Health Management of Complex Engineering Systems(2022) Lewis, Austin Drisko; Groth, Katrina M; Reliability Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md.)Complex engineering systems (CESes), such as nuclear power plants or manufacturing plants, are critical to a wide range of industries and utilities; as such, it is important to be able to monitor their system health and make informed decisions on maintenance and risk management practices. However, currently available system-level monitoring approaches either ignore complex dependencies in their probabilistic risk assessments (PRA) or are prognostics and health management (PHM) techniques intended for simpler systems. The gap in CES health management needs to be closed through the development of techniques and models built from a systematic integration of PHM and PRA (SIPPRA) approach that considers a system's causal factors and operational context when generating health assessments. The following dissertation describes a concentrated study that addresses one of the challenges facing SIPPRA: how to appropriately discretize a CES's operational timeline derived from multiple data streams to create discrete time-series data for use as model inputs over meaningful time periods. This research studies how different time scales and discretization approaches impact the performance of dynamic Bayesian Networks (DBNs), models that are increasingly used for causal-based inferences and system-level assessments, specifically built for SIPPRA health management. The impact of this research offers new insight into how to construct such DBNs to better support system-level health management for CESes.Item A BAYESIAN NETWORK PERSPECTIVE ON THE ELEMENTS OF A NUCLEAR POWER PLANT MULTI-UNIT SEISMIC PROBABILISTIC RISK ASSESSMENT(2021) DeJesus Segarra, Jonathan; Bensi, Michelle T.; Modarres, Mohammad; Mechanical Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md.)Nuclear power plants (NPPs) generated about 10% of the world’s electricity in 2020 and about 1/3 of the world’s low-carbon electricity production. Probabilistic risk assessments (PRAs) are used to estimate the risk posed by NPPs, generate insights related to strengths and vulnerabilities, and support risk-informed decisionmaking related to safety and reliability. While PRAs are typically carried out on a reactor-by-reactor basis, the Fukushima Dai-ichi accident highlighted the need to also consider multi-unit accidents. To properly characterize the risks of reactor core damage and subsequent radiation release at a multi-unit site, it is necessary to account for dependencies among reactors arising from the possibility that adverse conditions affect multiple units concurrently. For instance, the seismic hazard is one of the most critical threats to NPP structures, systems, and components (SSCs) because it affects their redundancy. Seismic PRAs are comprised of three elements: seismic hazard analysis, fragility evaluation, and systems analysis. This dissertation presents a Bayesian network (BN) perspective on the elements of a multi-unit seismic PRA (MUSPRA) by outlining a MUSPRA approach that accounts for the dependencies across NPP reactor units. BNs offer the following advantages: graphical representation that enables transparency and facilitates communicating modeling assumptions; efficiency in modeling complex dependencies; ability to accommodate differing probability distribution assumptions; and facilitating multi-directional inference, which allows for the efficient calculation of joint and conditional probability distributions for all random variables in the BN. The proposed MUSPRA approach considers the spatial variability of the ground motions (hazard analysis), dependent seismic performance of SSCs (fragility evaluation), and efficient BN modeling of systems (systems analysis). Considering the spatial variability of ground motions represents an improvement over the typical assumption that ground motions across a NPP site are perfectly correlated. The method to model dependent seismic performance of SSCs presented is an improvement over the current “perfectly dependent or independent” approach for dependent seismic performance and provides system failure probability results that comply with theoretical bounds. Accounting for these dependencies in a systematic manner makes the MUSPRA more realistic and, therefore, should provide confidence in its results (calculated metrics) and risk insights.Item TRITIATED NITROXIDE FOR BETAVOLTAIC CELL NUCLEAR BATTERY: 3D BETA FLUX MODELING, SYNTHESIS, STABILITY ANALYSIS, AND COATING TECHNIQUES(2019) Russo, John A; Bigio, David; Mechanical Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md.)Beta (β-) radioisotope energy sources, such as tritium (3H), have shown significant potential in satisfying the needs of a sensor-driven world. The limitations of current β- sources include: (i) low beta-flux power, (ii) intrinsic isotope leakage (instability) and (iii) beta self-absorption. The figure of merit is the β--flux power (dPβ/dS), where an optimal portion of incident beta particles penetrates the semiconductor depletion region. The goal of this research was to identify a compound to contain a beta emitter that can permit beta-flux power of at least 733nW/cm2 from one side, where it can be used for both planar and high aspect ratio microstructure technology (HARMST) transducers. Nitroxides were chosen because of previous demonstrated deuteration, ease of synthesis, diversity of structure, and pliability. A 6-membered nitroxide was prepared and tritiated with a specific activity of 103Ci/g. The product was stable after 256 days with only 2% tritium loss. A betavoltaic (βV) cell nuclear battery prototype was demonstrated with the tritiation of a 5-membered nitroxide stable in liquid and solid form and a specific activity (Am) of 635Ci/g, the highest recorded Am for an organic compound. A dPβ/dS of 51 nW/cm2 and 102 nW/cm2 were generated when dispensed on βV (4H-SiC) and PV (InGaP) cells. Improvements to increase the dPβ/dS closer to the theoretical limit were identified and demonstrated such as dispensing with different solvents to reduce evaporation time, and increasing solute (nitroxide) concentration in the dispensed solution. A βV cell nuclear battery model was developed, producing a blueprint on what nitroxide characteristics are required to maximize the dPβ/dS and electrical power density (P_(e,vol)). The percent error and percent difference were less than 6% compared to experimental data and other models. For the planar coupling configuration, increasing Am while increasing mass density increases P_(e,vol). Increasing the surface area interaction with the radioisotope and transducer increases the volume radioactivity (Vm), but does not always generate a higher β- source efficiency nor P_(e,vol) compared to the planar coupling configuration. The rectangular pillar array produced the highest 4.54 mW/cm3 of at the highest Vm where HARMST feature width and gaps are proportionally minimized at 100 nm wide.Item Evaluation of an Additively Manufactured Ferritic-Martensitic Steel for Advanced Nuclear Applications(2018) Vega, Daniel; Al-Sheikhly, Mohamad; Material Science and Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md.)A multidisciplinary investigation is presented in which the first known Additively Manufactured (AM) ferritic/martensitic (FM) steel alloys were evaluated for suitability as fast reactor structural components. As AM becomes more mainstream, it offers new possibilities in improving the design and cost of metal parts, especially those with weldability and workability limitations. However, questions remain about AM’s ability to reliably produce the types of high performance ferritic alloys that require carefully tailored microstructures. Laser-based AM produces heating patterns that interfere with the phase transformations from which wrought FM steels derive their ductility, high strength, and creep resistance. Additionally, study of irradiation effects on AM materials is immature. To address these questions, this dissertation presents an analysis of AM Grade 91 steel, an alloy with established pedigree in the nuclear and fossil fuel sectors, and whose ASME code case establishment was the first in a family of creep strength enhanced FM steels. Material from the first known successful AM build of Grade 91 steel was prepared, heat treated, analyzed using microstructural characterization techniques, subjected to a range of mechanical testing (to 600 °C), and irradiated up to 100 dpa with 5 MeV Fe2+ ions. Among the most salient findings were that i) AM material that was subjected to a prescribed normalization heat treatment developed a uniform microstructure and martensite fraction similar to wrought material, ii) normalized and tempered AM material had a similar distribution of carbide precipitates, but finer grain structure than wrought material, iii) AM material was slightly harder and less ductile than wrought material at room temperature, but at 300 °C and 600 °C, their mechanical strength/ductility were virtually the same, iv) AM heat treated material directly built and tested without heat treatments had an unpredictable and heterogeneous microstructure, but that when tensile tested, demonstrated extremely high strength and unexpectedly high ductility, especially at high temperatures, and iv) AM material showed less radiation-induced hardening, due to its fine grain structure. Indications are that AM Grade 91 steel may well be suitable for advanced nuclear applications, and additional research leading to a path forward for certification should be pursued.Item Development of Approaches to Common Cause Dependencies with Applications to Multi-Unit Nuclear Power Plant(2018) Zhou, Taotao; Modarres, Mohammad; Droguett, Enrique López; Mechanical Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md.)The term “common cause dependencies” encompasses the possible mechanisms that directly compromise components performances and ultimately cause degradation or failure of multiple components, referred to as common cause failure (CCF) events. The CCF events have been a major contributor to the risk posed by the nuclear power plants and considerable research efforts have been devoted to model the impacts of CCF based on historical observations and engineering judgment, referred to as CCF models. However, most current probabilistic risk assessment (PRA) studies are restricted to single reactor units and could not appropriately consider the common cause dependencies across reactor units. Recently, the common cause dependencies across reactor units have attracted a lot of attention, especially following the 2011 Fukushima accident in Japan that involved multiple reactor unit damages and radioactive source term releases. To gain an accurate view of a site's risk profile, a site-based risk metric representing the entire site rather than single reactor unit should be considered and evaluated through a multi-unit PRA (MUPRA). However, the multi-unit risk is neither formally nor adequately addressed in either the regulatory or the commercial nuclear environments and there are still gaps in the PRA methods to model such multi-unit events. In particular, external events, especially seismic events, are expected to be very important in the assessment of risks related to multi-unit nuclear plant sites. The objective of this dissertation is to develop three inter-related approaches to address important issues in both external events and internal events in the MUPRA. 1) Develop a general MUPRA framework to identify and characterize the multi-unit events, and ultimately to assess the risk profile of multi-unit sites. 2) Develop an improved approach to seismic MUPRA through identifying and addressing the issues in the current methods for seismic dependency modeling. The proposed approach can also be extended to address other external events involved in the MUPRA. 3) Develop a novel CCF model for components undergoing age-related degradation by superimposing the maintenance impacts on the component degradation evolutions inferred from condition monitoring data. This approach advances the state-of-the-art CCF analysis in general and assists in the studies of internal events of the MUPRA.Item Synthesis of Novel Co-Polymers Using Ionizing Radiation Grafting Methods for the Extraction of Uranium from Seawater(2017) Dietz, Travis Cameron; Al-Sheikhly, Mohamad I; Material Science and Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md.)The world’s oceans contain a relatively uniform uranium concentration of 3 μg/L. While this is an exceedingly small concentration, the quantity of uranium throughout the oceans is about 1000 times higher than the quantity in known terrestrial deposits. To take advantage of this immense resource, radiation grafting techniques were used to attach uranium-chelating monomers to durable polymer substrates. Three novel, uranium extracting co-polymer systems have been fabricated through this process and characterized. Three different compound classes were explored for their ability to extract uranium, specifically phosphates, oxalates, and azos. These classes displayed characteristics that provide advantages to the technology over state-of-the-art systems. For the phosphates and oxalates, monomers of these classes containing allyl groups were radiation grafted onto a polymer in a single step. For the azos, a chemical precursor containing a vinyl group was initially radiation grafted to a polymer. The azo compound was then chemically attached to the functionalized polymer surface. For effective seawater deployment, a polymer substrate was chosen as an inexpensive, reusable platform for extraction. While different fabric substrates were tested, high surface area (14 m^2/g) nylon 6 fabric was chosen for its durability and its capacity for radiation grafting. Direct and indirect radiation induced graft polymerization methods were used in this work. For direct grafting, the nylon 6 fabric was immersed in the monomer solution and irradiated. However, for indirect grafting, only the fabric was irradiated followed by the immediate introduction of the monomer solution. All of these experiments were conducted under anaerobic conditions to prevent the reaction of oxygen with the radiolytically-produced, carbon-centered free radicals. The grafted fabrics were characterized for attachment of the monomer and their ability to extract uranium. The degree of surface grafting was determined through attenuated total reflectance Fourier transform infrared spectroscopy, scanning electron microscopy, and energy dispersive X-ray spectroscopy, among other techniques. Electron paramagnetic resonance spectroscopy was used to determine radical decay kinetics in the polymer substrate. Pulse radiolysis was used to elucidate the polymerization reaction kinetics of certain monomers. These fabrics were then exposed to uranium-doped seawater solutions and the extraction capacities of the grafted materials were determined.Item MULTI-UNIT ACCIDENT CONTRIBUTIONS TO U.S. NUCLEAR REGULATORY COMMISSION QUANTITATIVE HEALTH OBJECTIVES: A SAFETY GOAL POLICY ANALYSIS USING MODELS FROM STATE-OF-THE-ART REACTOR CONSEQUENCE ANALYSES(2016) Hudson, Daniel Wayne; Modarres, Mohammad; Reliability Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md.)The U.S. Nuclear Regulatory Commission implemented a safety goal policy in response to the 1979 Three Mile Island accident. This policy addresses the question “How safe is safe enough?” by specifying quantitative health objectives (QHOs) for comparison with results from nuclear power plant (NPP) probabilistic risk analyses (PRAs) to determine whether proposed regulatory actions are justified based on potential safety benefit. Lessons learned from recent operating experience—including the 2011 Fukushima accident—indicate that accidents involving multiple units at a shared site can occur with non-negligible frequency. Yet risk contributions from such scenarios are excluded by policy from safety goal evaluations—even for the nearly 60% of U.S. NPP sites that include multiple units. This research develops and applies methods for estimating risk metrics for comparison with safety goal QHOs using models from state-of-the-art consequence analyses to evaluate the effect of including multi-unit accident risk contributions in safety goal evaluations.Item CHARACTERIZATION OF RADIATION DAMAGE TO A NOVEL PHOTONIC CRYSTAL SENSOR(2015) Grdanovska, Slavica; Briber, Robert M.; Koeth, Timothy W.; Nuclear Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md.)New methods of nuclear fuel and cladding characterization must be developed and implemented to enhance the safety and reliability of nuclear power plants. One class of such advanced methods is aimed at the characterization of fuel performance by performing minimally intrusive in-core, real time measurements on nuclear fuel on the nanometer scale. Nuclear power plants depend on instrumentation and control systems for monitoring, control and protection. Traditionally, methods for fuel characterization under irradiation are performed using a “cook and look” method. These methods are very expensive and labor-intensive since they require removal, inspection and return of irradiated samples for each measurement. Such fuel cladding inspection methods investigate oxide layer thickness, wear, dimensional changes, ovality, nuclear fuel growth and nuclear fuel defect identification. These methods are also not suitable for all commercial nuclear power applications as they are not always available to the operator when needed. Additionally, such techniques often provide limited data and may exacerbate the phenomena being investigated. This thesis investigates a novel, nanostructured sensor based on a photonic crystal design that is implemented in a nuclear reactor environment. The aim of this work is to produce an in-situ radiation-tolerant sensor capable of measuring the deformation of a nuclear material during nuclear reactor operations. The sensor was fabricated on the surface of nuclear reactor materials (specifically, steel and zirconium based alloys). Charged-particle and mixed-field irradiations were both performed on a newly-developed “pelletron” beamline at Idaho State University's Research and Innovation in Science and Engineering (RISE) complex and at the University of Maryland's 250 kW Training Reactor (MUTR). The sensors were irradiated to 6 different fluences (ranging from 1 to 100 dpa), followed by intensive characterization using focused ion beam (FIB), transmission electron microscopy (TEM) and scanning electron microscopy (SEM) to investigate the physical deformation and microstructural changes between different fluence levels, to provide high-resolution information regarding the material performance. Computer modeling (SRIM/TRIM) was employed to simulate damage to the sensor as well as to provide significant information concerning the penetration depth of the ions into the material.Item Radiation-Grafted Fabrics for the Extraction of Uranium from Seawater(2014) Tissot, Chanel; Briber, Robert; Nuclear Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md.)Much interest has been generated in extraction uranium from the ocean - the world's largest uranium reserve. This dissertation describes the development and seawater testing of a polymeric adsorbent for uranium based on radiation-induced grafting. Among all monomers and polymeric substrates tested, grafting of the monomer bis(2-methacryloxyethyl) phosphate (B2MP) onto Winged nylon fabric was determined to produce adsorbents of the highest degrees of grafting. Degree of grafting was optimized by irradiating at a range of dose rates and total absorbed doses and by varying monomer concentration, solvent, purging gas and radiation source. Both the University of Maryland's Co-60 gamma irradiator and 1-9 MeV pulsed LINAC were utilized. The grafted adsorbents were tested for uranium extraction capacity using a Uranium-233 radiotracer in synthetic seawater at natural (3.3 ppb) uranium concentrations. It was determined that adsorbents of degrees of grafting between 75 and 100% obtained the highest distribution coefficients for uranium. Kinetic studies revealed an increase in Uranium-233 concentration on the adsorbent over the course of 4 hours after which time a steady-state was reached. Correlation of this data with kinetic models indicated pseudo-second order kinetics, suggesting the rate-limiting adsorption mechanism as chemical complexation between Uranium-233 and the phosphate-containing adsorbent. Overall, the highest performing adsorbents obtained distribution coefficients of 1.2 × 104 mL/g and Uranium-233 loadings of 1.0 × 10-2 mg-U/g-adsorbent. These values were a result of performing the extractions at 3.3 Uranium-233, a concentration several orders of magnitude lower than those reported in similar studies. The chemical changes that occurred upon grafting were investigated with FTIR and Raman analysis of virgin, irradiated and B2MP-grafted Winged nylon. Characterization of the grafted adsorbents with SEM revealed a unique morphology for the grafted fabrics that has been attributed to the precipitation of homopolymer from the solvent during irradiation. SEM/EDS analysis of a grafted adsorbent contacted with Uranium-233-spiked synthetic seawater revealed the presence of several elements abundant in seawater, indicating that competition between uranium and other seawater ions is likely to limit the uranium uptake capacity of the adsorbent.