Department: Nuclear Engineering ![[clear]](close-x.png "Remove this limiter")
47 matches in the database.
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1.
Abejon Orzaez, Jorge.
Neutronics analysis of a modified Pebble Bed Advanced High Temperature Reactor.
Degree: MS, Nuclear Engineering, 2009, Ohio State University
► The objective of this research is to, based on the original design…
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▼ The objective of this research is to, based on the original design for the Pebble Bed Advanced High Temperature Reactor (PB-AHTR), develop an MCNPX model of the reactor core with the objective to attain criticality and to breed new fuel. A brief but complete description of a first approach to the PB-AHTR will be provided and a MCNPX model will be run in order to ascertain the difficulties of that configuration. On the second part, a modification of the original model will be evaluated and compared in order to resolve the difficulties encountered in the original design. Finally, in an effort to optimize the design, an evolutionary approach will be analyzed, based on the previous model, and conclusions will be attained
Advisors/Committee Members: Blue, Thomas.
Subjects: Engineering; Nuclear physics
Keywords: Neutronics, advanced high temperature reactor, criticality
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2.
Ablay, Gunyaz.
Sliding Mode Approaches for Robust Control, State Estimation, Secure Communication, and Fault Diagnosis in Nuclear Systems.
Degree: PhD, Nuclear Engineering, 2012, Ohio State University
► Using traditional control methods for controller design, parameter estimation and fault diagnosis…
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▼ Using traditional control methods for controller design, parameter estimation and fault diagnosis may lead to poor results with nuclear systems in practice because of approximations and uncertainties in the system models used, possibly resulting in unexpected plant unavailability. This experience has led to an interest in development of robust control, estimation and fault diagnosis methods. One particularly robust approach is the sliding mode control methodology. Sliding mode approaches have been of great interest and importance in industry and engineering in the recent decades due to their potential for producing economic, safe and reliable designs. In order to utilize these advantages, sliding mode approaches are implemented for robust control, state estimation, secure communication and fault diagnosis in nuclear plant systems. In addition, a sliding mode output observer is developed for fault diagnosis in dynamical systems. To validate the effectiveness of the methodologies, several nuclear plant system models are considered for applications, including point reactor kinetics, xenon concentration dynamics, an uncertain pressurizer model, a U-tube steam generator model and a coupled nonlinear nuclear reactor model.
Advisors/Committee Members: Aldemir, Tunc.
Subjects: Nuclear Engineering
Keywords: nuclear reactors; sliding modes; fault diagnosis; robust control; steam generator; reactor kinetics; observers; chaos; secure communication
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3.
Arcilesi, David J. Jr.
Developmental Analysis and Design of a Scaled-down Test Facility for a VHTR Air-ingress Accident.
Degree: MS, Nuclear Engineering, 2012, Ohio State University
► A critical event in the safety analysis of the Very High-temperature Gas-cooled…
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▼ A critical event in the safety analysis of the Very High-temperature Gas-cooled Reactor (VHTR) is a loss-of-coolant accident (LOCA). This accident is initiated, in its worst case scenario, by a double-ended guillotine break of the hot duct, which leads to a rapid reactor depressurization. In a VHTR, the reactor vessel is located within a reactor cavity that is filled with air during normal operating conditions. During a LOCA, an air-helium mixture may enter the reactor vessel following a reactor vessel depressurization. Since air chemically reacts with high-temperature graphite, this could lead to damage of core-bottom and in-core graphite structures as well as core heat-up, toxic gas release, and failure of the structural integrity of the system unless mitigating actions are taken. Therefore, it is imperative to understand the dominant mechanism(s) in the air-ingress process so that mitigating measures can be considered for VHTR designs. Early studies postulated that the dominant mechanism of air ingress is molecular diffusion. In general, however, molecular diffusion is a slow process, and recent studies show that the air-ingress process could be initially controlled by density-driven stratified flow of hot helium and a relatively cool air-helium mixture in the hot duct. If density-driven stratified flow initially dominates, earlier onset of natural circulation within the core would occur. This would lead to an earlier onset of oxidation of internal graphite structures and, most likely, at a more rapid rate. Thus, it is important to understand both of these air ingress mechanisms in a VHTR. These mechanisms may be important at different times for different scenarios, specifically breaks of varying size, orientation, shape, and location. Since no experimental data are readily available to understand the phenomena and determine which mechanism will dominate for various break conditions, there’s a need to design and construct a scaled-down experimental test facility to generate data. In this thesis, the scaling analysis, the developmental analysis and design of a scaled-down air-ingress accident test facility will be given. As part of the developmental analysis, the non-dimensional Froude number was preserved in establishing hydraulic similarity. On average, the non-dimensional resistance number of the scaled-down facility deviates 2.81% in terms of relative accuracy from the non-dimensional resistance number of the prototypic design. A 1/8th geometric scale is utilized for the entire geometry except for the hot duct length, support column pitch and support column diameter. The exceptions to the 1/8th geometric scale are to avoid large distortion of the loop pressure loss distribution (modified hot duct length) and to preserve the non-dimensional Froude number (modified support column diameter and pitch). A heat transfer characterization of the lower plenum of the prototypic and scaled-down system was performed. The characterization focused on the support columns which are the principal heat source in the lower plenum during an air-ingress accident scenario. This analysis shows that a lumped capacitance approximation for the support columns is valid. Also, the analysis determines an operational heater power ( = 125 W) for shell/heater rods in the scaled-down system so that the rod surface temperature and the rod average radial heat flux ( = 0 W) can be preserved from the prototypic case. In addition, a containment free volume (V = 1 m3) was determined to house the scaled-down facility. With the containment free volume known, initial vessel pressures to preserve the air-to-helium mole ratio (P = 40 psig) and mixed mean temperature (P = 34.2 psig) of the prototype case were calculated. Finally, vessel design drawings and instrumentation are given
Advisors/Committee Members: Christensen, Richard.
Subjects: Nuclear Engineering
Keywords: VHTR; Air-ingress Accident; Nuclear Engineering; high temperature gas-cooled reactor
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4.
Arndt, Steven Andrew.
Methods and Strategies for Future Reactor Safety Goals.
Degree: PhD, Nuclear Engineering, 2010, Ohio State University
► The NRC, in its safety goals policy statement, has provided general qualitative…
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▼ The NRC, in its safety goals policy statement, has provided general qualitative safety goals and basic quantitative health objectives (QHOs) for nuclear reactors in the United States. Risk metrics such as core damage frequency (CDF) and large early release frequency (LERF) have been used as surrogates for the QHOs. A number of issues have been raised including how risk from multiple reactors at a site should be combined for evaluation, how the combination of a new and old reactor at the same site should be evaluated, what the criteria for evaluating new reactors should be, and whether new reactors should be required to be safer than current generation reactors. As part of the development and application of the NRC safety goal policy statement the Commissioners laid out the expectations for the safety of a nuclear power plant but did not address the risk associated with current multi-unit sites, potential modular reactor sites, and hybrid sites that could contain current generation reactors, new passive reactors, and/or modular reactors. This dissertation examines potential approaches to updating the safety goals that include the establishment of new quantitative safety goal associated with the comparative risk of generating electricity by viable competing technologies and modifications of the goals to account for multi-plant reactor sites, and issues associated with the use of safety goals in both initial licensing and operational decision making. This research develops a new quantitative health objective that uses a comparable benefit risk metric based on the life-cycle risk of the construction, operation and decommissioning of comparable non-nuclear electric generation facility, as well as the risks associated with the mining, transportation. This dissertation also evaluates the effects of using various methods for aggregating site risk as a safety metric. Additionally, a number of important assumptions inherent in the current safety goals, including the effect of other potential negative societal effects such as the generation of greenhouse gases (e.g., carbon dioxide) have on the risk of electric power production and their effects on the setting of safety goals is explored. Finally the role risk perception should play in establishing safety goals, has been explored. To complete this evaluation a new method to analytically compare alternative technologies of generating electricity was developed, including development of a new way to evaluate risk perception and a new method was developed for evaluating the risk at multiple units on a single site. To test these modifications to the safety goals a number of possible reactor designs and configurations were evaluated using these new proposed safety goals to determine the goals’ usefulness and utility. The results of the analysis showed that the modifications provide measures that more closely evaluate the potential risk to the public from the operation of nuclear power plants that the current safety goals, while still providing a straight forward process for assessment of reactor design and operation.
Advisors/Committee Members: Denning, Prof. Richard.
Subjects: Mechanical engineering
Keywords: Reactor Safety; Safety Goals
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5.
Baas, Larry Brandon.
Feasibility Study of Concept Designs for Photonic Radiation Detection.
Degree: MS, Nuclear Engineering, 2009, Ohio State University
► The objective of this research is to do a feasibility study on…
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▼ The objective of this research is to do a feasibility study on a photonic-based semiconductor concept design for a radiation detector. The research was funded by the National Science Foundation in conjunction with the Department of Homeland Security. It was a cooperative effort between the Nuclear Engineering Program and the Electrical and Computer Engineering Department at Ohio State University. There are two concept device configurations considered in this study, the critical angle setup and the Fabry-Perot setup. In addition to being able to detect radiation, the device should be portable, durable and user friendly for field use. Its initial purpose is to be used to detect gamma radiation, but it is not limited to this purpose. Semiconductors have already been successfully used in highly sensitive radiation detectors (e.g. germanium detectors). The perceived advantage is that photonics can possibly be used to make a sensitive detector that can operate while at room temperature and be only slightly affected by electromagnetic interference due to its operating nature. Two semiconductor materials, silicon and cadmium telluride, were analyzed in detail, while others were considered as possibilities for later research. Following the development of an analytical model, experimental data were taken to validate the model. The experiments were performed for silicon because it is the least expensive of the candidate semiconductor materials.
Advisors/Committee Members: Blue, Thomas.
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6.
Blake, Bryan P.
Initial Testing of Single-Mode Optical Fibers Interrogated with an Optical Backscatter Reflectometer at High Temperatures and in Radiation Environments for Advanced Instrumentation in Nuclear Reactors.
Degree: MS, Nuclear Engineering, 2012, Ohio State University
► Although optical fibers have been used in the telecommunications industry for decades,…
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▼ Although optical fibers have been used in the telecommunications industry for decades, the extreme temperature and radiation environments of next generation nuclear reactors has recently led to the exploration of optical fibers for advanced instrumentation and control. Optical fibers appear to be well suited for high temperature nuclear applications because they are immune to electromagnetic interference, can withstand harsh environments, and are capable of performing distributed sensing. Previous work with optical fibers has focused on the transmission of light. This thesis presents an alternative interferometric technique that measures the Rayleigh backscatter as a function of position using Luna Technologies’ Optical Backscatter Reflectometer (OBR). The OBR has the capability to perform distributed temperature and strain sensing. However, this work investigated the optical backscattered signal as a function of position when the Corning SMF-28e+ single-mode optical fiber was exposed to high temperatures and radiation environments. The work has created a large set of OBR data that was analyzed to characterize single-mode fiber interrogated by the OBR at high temperatures and in radiation environments by monitoring the average fiber Amplitude (dB). Three experiments were performed: a 100 hour thermal-only experiment up to 1000°C, a 73 hour gamma irradiation experiment in the cobalt-60 irradiator at The Ohio State University Nuclear Reactor Laboratory up to 400°C, and a 450 hour reactor irradiation experiment at The Ohio State University Research Reactor up to 1000°C. This work has proven that optical amplitude measurements can be performed with the OBR at high temperatures in radiation environments. A temporary temperature effect was seen where the average fiber Amplitude (dB) increased with increasing temperature and decreased with decreasing temperature. Slight permanent damage was seen in the thermal-only and reactor irradiation experiments as a decrease in average fiber Amplitude (dB) due to high temperatures. The maximum decrease in average fiber Amplitude (dB) for the thermal-only and reactor irradiation experiments was less than 1 dB and 6 dB, respectively. Even at high temperatures in a reactor irradiation environment, light propagates to the heated and irradiated region of the test fiber and is scattered back with a large enough fraction to measure a signal that is distinguishable from the noise floor. As a next step, the experimental data acquired in these experiments will be analyzed with Luna Technologies’ distributed sensing software to extract temperature information to determine if the OBR can function as a distributed sensor at high temperatures in radiation environments.
Advisors/Committee Members: Blue, Thomas.
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7.
Bledsoe, Keith C.
Inverse Methods for Radiation Transport.
Degree: PhD, Nuclear Engineering, 2009, Ohio State University
► Implicit optimization methods for solving the inverse transport problems of interface location…
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▼ Implicit optimization methods for solving the inverse transport problems of interface location identification, source isotope weight fraction identification, shield material identification, and material mass density identification are explored. Among these optimization methods are the Schwinger inverse method, Levenberg-Marquardt method, and evolutionary algorithms. Inverse problems are studied in one-dimensional spherical and two-dimensional cylindrical geometries. The scalar fluxes of unscattered gamma-ray lines, leakages of neutron-induced gamma-ray lines, and/or neutron multiplication in the system are assumed to be measured. Each optimization method is studied on numerical test problems in which the measured data is simulated using the same deterministic transport code used in the optimization process (assuming perfectly consistent measurements) and using a Monte Carlo code (assuming less-consistent, more realistic measurements). The Schwinger inverse method and Levenberg-Marquardt methods are found to be successful for problems with relatively few (i.e. 4 or fewer) unknown parameters, with the former being the best for unknown isotope problems and the latter being more adept at interface location, unknown material mass density, and mixed parameter problems. A study of a variety of evolutionary algorithms indicates that the differential evolution method is the best for inverse transport problems, and outperforms the Levenberg-Marquardt method on problems with large numbers of unknowns. An algorithm created by combining different variants of the differential evolution method is shown to be highly successful on spherical problems with unscattered gamma-ray lines, while a basic differential evolution approach is more useful for problems with scattering and in cylindrical geometries. A hybrid differential evolution/Levenberg-Marquardt algorithm also was found to show promise for fast and robust solution of inverse problems.
Advisors/Committee Members: Aldemir, Tunc.
Subjects: Engineering
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8.
Bratton, Isaac John.
Modeling and Validation of the Fuel Depletion and Burnup of the OSU Research Reactor Using MCNPX/CINDER'90.
Degree: MS, Nuclear Engineering, 2012, Ohio State University
► The purpose of this study was to provide an accurate nuclide inventory,…
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▼ The purpose of this study was to provide an accurate nuclide inventory, depletion and burnup of The Ohio State University Research Reactor (OSURR) Low Enriched Uranium (LEU) core using Monte Carlo techniques. This work will allow for a better understanding of the OSURR, thereby improving the risk-informed regulatory process, improving computer model predictive capability, and increasing understanding of the remaining fuel and reactor lifetime. This thesis used two separate methods for depletion analysis: a simple, relatively quick method, and a more detailed, computationally expensive method. The simple method calculated the consumption of 235U from the operational history and used a consumption rate of 1.23 g 235U /MWd. The depletion distribution was set proportional to the Monte Carlo N-Particle (MCNP) calculated fission reaction rate distribution. The second method tracked the evolution of all actinides and non-actinides produced using MCNPX and CINDER’90. MCNPX calculates the flux spectrum and reaction rates, which are then passed to CINDER’90 for depletion calculations and material nuclide inventory updates. Both methods used an updated MCNP model to track the depletion distribution throughout the OSURR core. The models were benchmarked against operational data from the reactor. The results of both methods were very comparable. The total 235U depletion, over the LEU core lifetime, was 65.83 grams and 66.61 grams for methods 1 and 2, respectively. It was found that approximately five grams of 239Pu was produced, contributing to the reactor fuel and power. Results showed that, while there was some variation of burnup between the fuel elements based on core position, the overall effect was negligible on effective fuel management of the core, and was well below burnup limits for structural integrity. The principal conclusion was that the fuel depletion is relatively small; however, due to the insertion of experimental facilities additional fuel assemblies have been added to maintain constant and sufficient excess reactivity. This thesis was an important step in initiating a fuel management program for the OSURR. In addition, the updated MCNP model of the OSURR will provide accurate calculations and better predictive capability of experimental results. The OSURR fuel management program will continue to follow core nuclide inventory. Future operation predictions can be used to predict core performance, fuel lifetime and optimize the core pattern.
Advisors/Committee Members: Blue, Thomas.
Subjects: Nuclear Engineering
Keywords: fuel management; depletion; MCNPX; CINDER90; burnup; research reactor
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9.
Breuning, David.
A Low-Level Radioactive Waste Management Program for Activated Waste from he GE PETrace™ Cyclotron.
Degree: MS, Nuclear Engineering, 2010, Ohio State University
► Preventative Maintenance and Operational Use of the 16.5 MeV GE PETrace™ Cyclotron,…
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▼ Preventative Maintenance and Operational Use of the 16.5 MeV GE PETrace™ Cyclotron, used for the production of 18F and other radioisotope labeled pharmaceuticals, produces long-lived activated waste products which must be classified by radionuclide and activity before disposal as Low-Level Radioactive Waste (LLRW). Large volumes of this waste can be created in a busy cyclotron facility, but full gamma spectrometric analysis of all of the waste is expensive, and excessive handling of the waste is counter to the As Low As Reasonably Achievable (ALARA) concept. However, analysis of representative waste components allows them to be classified into categories in a manner such that all components in any category contain similar radioisotopic content. Radioisotopic activity and exposure rate measurements of representative samples of each of these categories can be used as standards to create conversion factors for each radionuclide in each category in units of activity per unit exposure rate. Subsequent future waste assumes that similar processes produce similar mixtures of radioisotopes, allowing this future waste to placed in the same category in which the standard was placed. This thesis assesses with what accuracy an estimate of the maximum possible isotopic activity of the waste can be determined by taking exposure rate measurements of future waste and multiplying the measured exposure rates by each conversion factor for each radionuclide for the category. The results of the assessment are that for Niobium and Silver Targets, Vacuum Tank and Ion Source Rags, Delivery Lines and other Delivery System Components, Chimneys and Pullers, Vacuum Tank Rods and Seals, Collimators and Baffles, and Extraction Foils and Fittings an estimate of the maximum possible isotopic activity of the waste, can be determined with sufficient accuracy to allow it to be disposed of as LLRW
Advisors/Committee Members: Blue, Thomas.
Subjects: Engineering; Nuclear physics; Pharmaceuticals; Radiation
Keywords: radioactive waste; PETrace; activated waste; cyclotron
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10.
Brunett, Acacia Joann.
A Methodology for Analyzing the Consequences of Accidents in Sodium-Cooled Fast Reactors.
Degree: MS, Nuclear Engineering, 2010, Ohio State University
► The objective of this research is to develop methods to analyze the…
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▼ The objective of this research is to develop methods to analyze the offsite consequences of specific sodium fast reactor (SFR) licensing basis events as used to characterize the safety of a nuclear plant design in a license application. An algorithm has been developed which assesses pool heatup, containment load, radionuclide transport, and release from containment during accident scenarios. To analyze temperature transients, the SFR pool has been divided into two, well-mixed regions separated by a metal divider referred to as a redan. Heatup due to fission power or decay heat in the core and heat removal by either the Intermediate Heat Exchanger (IHX) or passive heat removal system drive the energy balance. A reactor kinetics model is also included in this analysis. Radionuclide transport from the cover gas region to containment, as well as the heat and pressure load on containment are analyzed using MELCOR. Subsequent releases from containment to the environment are then analyzed using WinMACCS or Regulatory Guide 1.145 to calculate offsite consequences. Both WinMACCS and Regulatory Guide 1.145 employ a Gaussian dispersion model and meteorological data from an Eastern U.S. site. Several accident scenarios are examined in the radionuclide release and transport analysis: varying sizes of sodium spills inside containment, arrested melt scenarios, an energetic recriticality event, core uncovery and a case of sustained sodium vaporization. The scenarios will be analyzed for both containment intact and containment failed, where the containment has a design basis leak rate of 1.0 v/o per day. Resulting offsite consequences are then compared with the NRC’S frequency-consequence (F-C) curve in the Technology Neutral Framework (TNF) for compliance. Calculations are also performed to examine the implications of satisfying the F-C curve and its relationship to satisfaction of the NRC’s Quantitative Health Objectives.
Advisors/Committee Members: Denning, Richard.
Subjects: Engineering
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11.
Bucknor, Matthew David.
Study of In-Core Flow Blockage by Insulation Debris.
Degree: MS, Nuclear Engineering, 2009, Ohio State University
► The objective of this research is to determine the thickness of a…
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▼ The objective of this research is to determine the thickness of a debris bed comprised of insulation material that would lead to a peak cladding temperature of 2200°F, which is a thermal limit specified in the Federal Code of Regulations for Nuclear Power Plants, following a loss of coolant accident. This study considers the case of a cold-leg break which leads to a loss of reactor coolant. In this accident, fiberglass debris is generated from the breakup of thermal insulation during the blowdown phase in the accident. The material is then transported to the sump where it may be pumped into the core where it is possible to build up a debris bed at spacer grid locations. The debris bed will restrict coolant flow and therefore lead to deteriorated heat transfer from the cladding to the coolant. A two-dimensional heat transfer analysis is performed to determine the critical length of blanketed region beyond which axial conduction in the pin is no longer adequate to prevent the peak cladding temperature from exceeding the regulatory limit. The feasibility of creating a debris bed thickness greater than the critical size is also evaluated. The amount of insulation debris potentially available to the core region is compared with the amount required to create a debris bed of the critical thickness.
Advisors/Committee Members: Denning, Richard.
Subjects: Mechanical engineering
Keywords: in-core blockage; cold leg break; nukon; core blockage
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12.
Chenkovich, Robert Jeremy.
Refinement and Validation of Existing Computer Models of the OSU Research Reactor using Activation Analysis and Spectral Unfolding Codes.
Degree: MS, Nuclear Engineering, 2008, Ohio State University
► The objective of this work is to provide more accurate neutronic models…
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▼ The objective of this work is to provide more accurate neutronic models of the OSU research reactor and also to provide a more complete characterization of the energy-dependent neutron flux spectrum present at the various experimental facilities of the reactor. This work will allow for a better predictive capability, so that experimental results can be better anticipated. Two models are analyzed, one with a very detailed core geometry, the other with a more homogenized core. The two models were refined and made to be more accurate in various ways. Also, experimental benchmarking of these computer models is included, specifically using material activation experiments and spectral deconvolution (or unfolding) codes. In these programs, the wires' post-irradiation activities, their response functions, and an initial guess of the energy-dependent neutron flux spectrum are input into an iterative process, which outputs a energy-dependent neutron flux spectrum. The experimental and predicted energy-dependent neutron flux spectra are then compared to illustrate the differences between models.
Advisors/Committee Members: Blue, Thomas.
Subjects: Engineering
Keywords: neutron flux; spectral analysis; spectral unfolding; activation; MCNP
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13.
Doup, Benjamin.
Experimental Investigation of Flow Structure Development in Air-water Two-phase Flows.
Degree: MS, Nuclear Engineering, 2012, Ohio State University
► The objective of this research is to help in the development of…
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▼ The objective of this research is to help in the development of a dynamic flow regime transition model by providing accurate experimental two-phase flow data near flow regime transitions. The experimental data will be used to help in understanding the transitions better and to benchmark/validate the dynamic flow regime transition model. The experiments are performed using a vertical cylindrical air-water two-phase flow loop that has maximum superficial gas and liquid velocities of 6.0 and 1.7 m/s, respectively. Measurements are carried out at three different axial location z/D = 10, 32, and 54 using a high-speed video camera, impedance void-meter signals, and four-sensor conductivity probes to capture the flow structure. The four-sensor conductivity probes provide 14 local radial measurements of the void fraction, gas velocity, interfacial area concentration, and Sauter mean bubble diameter at each measurement location. To validate the experimental data, the superficial gas velocity found by an area integral of the product of the local gas velocity and void fraction, that are measured using four-sensor conductivity probes, is compared with the superficial gas velocity found using rotameters and pressure measurements and the difference is generally below ±20%. The area-averaged void fractions and the void-weighted area-averaged gas velocities from the four-sensor conductivity probes are compared with their corresponding drift-flux model predicted values and the difference is generally below ±15%. The high-speed images are used to classify the flow conditions into flow regimes and this classification is used to compare the performance of a dynamic flow regime transition model that has been introduced in Wang et al. [1] and a flow regime map based on transition criteria outline in Mishima and Ishii [2]. From this comparison, it appears that the cap-bubbly to slug flow transition criteria and the slug to churn-turbulent flow transition criteria of the dynamic flow regime transition model need to be improved. The current test matrix only has two flow conditions where churn-turbulent flows are observed, more of these flow conditions need to be added to the test matrix to have a more complete benchmark of this dynamic flow regime transition model.
Advisors/Committee Members: Sun, Xiaodong.
Subjects: Nuclear Engineering
Keywords: two-phase flow, flow structure, flow development
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14.
Figley, Justin T.
Numerical Modeling and Performance Analysis of Printed Circuit Heat Exchanger for Very High-Temperature Reactors.
Degree: MS, Nuclear Engineering, 2009, Ohio State University
► Very High Temperature Reactors (VHTRs) operate at high temperatures (1,173-1,223 K) and…
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▼ Very High Temperature Reactors (VHTRs) operate at high temperatures (1,173-1,223 K) and require intermediate heat exchangers to transfer thermal energy to a hydrogen production plant or power conversion system. A promising plate-type compact heat exchanger for these applications is the Printed Circuit Heat Exchanger (PCHE). The objective of this study is to numerically model an Alloy 617 PCHE core with Helium as the working fluid using Fluent™ computational fluid dynamics software. The PCHE dimensions and operating conditions are those of a high-temperature helium test facility under construction at The Ohio State University. The test conditions considered are based upon the nominal design conditions of the test facility: operating pressure up to 3 MPa, mass flow rates of 10 to 80 kg/h, and hot and cold side inlet temperatures of 1,173 and 813 K, respectively. These operating conditions correspond to laminar and laminar-to-turbulent transitional flows within the fluid passages of the PCHEs being fabricated and modeled. The overall heat transfer coefficient ranges from 563-1697 W/m2K. The maximum effectiveness achieved is 85%. The maximum pressure drop of this PCHE is found to be approximately 1.5% of the operating pressure. The thermal duty of the heat exchanger ranges from 4.45 to 28.73 kW. The critical Reynolds number is found to be approximately 2800 for the semicircular channel as opposed to 2300 for a circular channel. CFD simulations carried out for laminar flow operating conditions are within good agreement with the predictions made using published correlations and empirical data. CFD simulations carried out for low Reynolds number laminar-to-turbulent transition cases are not accurately predicted by the correlations recommended in the published literature.
Advisors/Committee Members: Sun, Xiaodong.
Subjects: Energy; Engineering; Mechanical engineering
Keywords: Intermediate Heat Exchanger; Thermal Hydraulics; VHTR; Printer Circuit Heat Exchanger; PCHE
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15.
Flanders, Justin M.
Thermal Transport and Heat Exchanger Design for the Space Molten Salt Reactor Concept.
Degree: MS, Nuclear Engineering, 2012, Ohio State University
► Surface power and nuclear electric propulsion in space necessitate the development of…
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▼ Surface power and nuclear electric propulsion in space necessitate the development of high energy density, long term continuous power sources. Research at The Ohio State University under the NASA Ralph Steckler Space Grant Colonization Research and Technology Development Opportunity has identified molten salt reactors (MSRs) as a potentially appealing technology for high power, high temperature space fission systems. This thesis examines component specific design related to thermal transport in an attempt to further establish the feasibility of MSRs in space. Specifically, the optimization of the associated heat exchangers for a Brayton power cycle is discussed in detail. In addition, power cycle and secondary coolant selection, material selection, and general MSR design considerations are discussed. The primary heat exchanger was optimized to maximize the margin to super prompt critical (MSPC), with a final MSPC equal to 513 pcm. The secondary and tertiary heat exchangers were optimized to minimize helium pressure drop, with a combined pressure loss of 2.62 kPa obtained.
Advisors/Committee Members: Blue, Thomas.
Subjects: Nuclear Engineering
Keywords: molten salt; molten salt reactors; NASA; space nuclear; Brayton; heat exchanger; offset strip fin
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16.
Glosup, Richard Edwin.
Characterization of the High-Temperature Helium Facility in the Thermal Hydraulics Laboratory.
Degree: MS, Nuclear Engineering, 2011, Ohio State University
► The U.S. Department of Energy’s (DOE) Office of Nuclear Energy is actively…
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▼ The U.S. Department of Energy’s (DOE) Office of Nuclear Energy is actively pursuing design research related to Generation IV nuclear reactors for near term deployment as commercial nuclear facilities. High-temperature reactors are expected to provide collateral process heat for industrial applications that require high-temperatures during manufacturing processes. For the DOE to complete its research and development goals, all related components, instrumentation, and materials require assessment in high-temperature environments. The High-Temperature Helium Facility (HTHF) located in the Thermal Hydraulics Laboratory at The Ohio State University was designed to provide a laboratory environment that will be integral to completion of research in these areas. Intended to complement research and development projects that will help the industry reach the objectives set forth by the Generation IV International Forum for the development of next generation, high-temperature gas cooled reactors, construction of the HTHF has been completed. This document briefly reviews the HTHF’s layout and operational characteristics. Specifically, a detailed list of the facility’s instrumentation and an uncertainty analysis for each component is presented along with a description of the procedures developed to calibrate the thermocouples monitoring the HTHF and their application in the test environment. Calibration results are presented in conjunction with a brief review of thermocouple design and functionality. The calibration results clearly prove that the K-type thermocouples are within their specified tolerances. Calibration, application, and uncertainties of three Venturi meters are detailed herein. It has been observed that one of the three Venturi meters is not functioning according to its calibration. Heat losses related to the heaters, piping, test sections, and instrumentation ports are fully documented and do not show significant deviation from the losses specified in the facility’s design documentation and analyses. Pressure drop across the facility corresponds to known correlations and is nominally between 10 and 20 psi (0.07 and 0.14 MPa). To date the facility’s performance includes a maximum temperature, pressure and mass flow rate of 640 °C, 400 psig (2.8 MPa), and 50 kg/hr, respectively, using helium as the cooling agent. Following additional testing and possible minor modifications, the facility is expected to provide a testing environment that achieves 800 °C at 435 psig (3 MPa) and 40 kg/hr, which more closely approximates the temperature requirements for high-temperature gas cooled reactors. The HTHF is currently operational with the ability to accommodate various components for testing in a high-temperature environment.
Advisors/Committee Members: Sun, Xiaodong.
Subjects: Energy; Engineering; Nuclear Engineering
Keywords: High-temperature; characterization
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17.
Goodenow, Debra A. B.S.
Characterization of Effects of Mixed Neutron/Gamma Irradiation on NASA Glenn SiC Piezoresistive Pressure Sensors.
Degree: MS, Nuclear Engineering, 2008, Ohio State University
► SiC piezoresistive pressure sensors were developed by NASA Glenn Research Center GRC)…
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▼ SiC piezoresistive pressure sensors were developed by NASA Glenn Research Center GRC) for operation in high temperature environments. SiC's wide band gap and good radiation resistance make SiC piezoresistive pressure sensors potentially suitable for use in high radiation environments. In this thesis the degradation of SiC piezoresistive pressure sensors in a Wheatstone bridge configuration is evaluated in gamma and mixed neutron/gamma radiation fields. The performance of a Si piezoresistive pressure sensor in a Wheatstone bridge configuration was also evaluated in the same radiation fields for comparison. The degradation of the performance of the Si and SiC pressure sensors was evaluated by comparing the voltage response of the sensors to pressure before, during and after irradiation. There was noticeable degradation to the voltage response of the sensors. The change in the performance of the SiC pressure sensor due to radiation was compared to measurements of the effects of temperature on the sensor performance as determined by NASA GRC. The effect of radiation on the SiC sensor's performance was very similar to the effect of temperature on the SiC sensor's performance.
Advisors/Committee Members: Blue, Thomas.
Subjects: Aerospace materials; Electrical engineering; Engineering; Mechanical engineering; Nuclear physics; Radiation
Keywords: SiC, radiation, temperature, wheatstone bridge,piezoresistive
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18.
Grabaskas, David.
Analysis of Transient Overpower Scenarios in Sodium Fast Reactors.
Degree: MS, Nuclear Engineering, 2010, Ohio State University
► The objective of this research was the demonstration of a methodology to…
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▼ The objective of this research was the demonstration of a methodology to handle epistemic and aleatory uncertainties within an unprotected transient overpower (TOP) scenario in a sodium fast reactor. This study included a comparison of the risk of TOP scenarios to safety limits that are being considered by the U.S. Nuclear Regulatory Commission for future reactor designs. An analysis of the relative importance of uncertainties was also made. The results of this experiment demonstrated that the core damage frequency for an unprotected transient overpower scenario fell well below the proposed safety limits, even without taking credit for the reliability of the reactor protection system. The control rod driveline feedback and Doppler feedback mechanisms were found to be the most important epistemic uncertainties.
Advisors/Committee Members: Denning, Richard.
Subjects: Engineering; Nuclear physics
Keywords: Sodium Fast Reactor; Unprotected Transient Overpower; Uncertainty Analysis; Aleatory Uncertainty; Epistemic Uncertainty
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19.
Grabaskas, David.
Efficient Approaches to the Treatment of Uncertainty in Satisfying Regulatory Limits.
Degree: PhD, Nuclear Engineering, 2012, Ohio State University
► Utilities operating nuclear power plants in the United States are required to…
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▼ Utilities operating nuclear power plants in the United States are required to demonstrate that their plants comply with the safety requirements set by the U.S. Nuclear Regulatory Commission (NRC). How to show adherence to these limits through the use of computer code surrogates is not always straightforward, and different techniques have been proposed and approved by the regulator. The issue of compliance with regulatory limits is examined by rephrasing the problem in terms of hypothesis testing. By using this more rigorous framework, guidance is proposed to choose techniques to increase the probability of arriving at the correct conclusion of the analysis. The findings of this study show that the most straightforward way to achieve this goal is to reduce the variance of the output result of the computer code experiments. By analyzing different variance reduction techniques, and different methods of satisfying the NRC’s requirements, recommendations can be made about the best-practices, that would result in a more accurate and precise result. This study began with an investigation into the point estimate of the 0.95-quantile using traditional sampling methods, and new orthogonal designs. From there, new work on how to establish confidence intervals for the outputs of experiments designed using variance reduction techniques was compared to current, regulator-approved methods. Lastly, a more direct interpretation of the regulator’s probability requirement was used, and confidence intervals were established for the probability of exceeding a safety limit. From there, efforts were made at combining methods, in order to take advantage of positive aspects of different techniques. The results of this analysis show that these variance reduction techniques can provide a more accurate and precise result compared to current methods. This means an increased probability of arriving at the correct conclusion, and a more accurate characterization of the risk associated with events. While several of these methods are asymptotic in nature, which presents potential drawbacks, issues of convergence appear to be outweighed by the reduction in variance, and improvement of the information contained in the results. Using this knowledge, recommendations were made about the applicability of these methods in the field of reactor safety, and about future regulatory limits and their implications.
Advisors/Committee Members: Aldemir, Tunc.
Subjects: Nuclear Engineering; Statistics
Keywords: Latin Hypercube Sampling; Orthogonal Arrays; Orthogonal Latin Hypercubes; Monte Carlo Sampling; Confidence Intervals; Quantiles; Regulatory Limits; Nuclear Regulatory Commission; MELCOR; Probability; Safety Limits; Safety Goals; Margin; Hypothesis Testing
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20.
Hakobyan, Aram P.
Severe accident analysis using dynamic accident progression event trees.
Degree: PhD, Nuclear Engineering, 2006, Ohio State University
► In present, the development and analysis of Accident Progression Event Trees (APETs)…
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▼ In present, the development and analysis of Accident Progression Event Trees (APETs) are performed in a manner that is computationally time consuming, difficult to reproduce and also can be phenomenologically inconsistent. One of the principal deficiencies lies in the static nature of conventional APETs. In the conventional event tree techniques, the sequence of events is pre-determined in a fixed order based on the expert judgments. The main objective of this PhD dissertation was to develop a software tool (ADAPT) for automated APET generation using the concept of dynamic event trees. As implied by the name, in dynamic event trees the order and timing of events are determined by the progression of the accident. The tool determines the branching times from a severe accident analysis code based on user specified criteria for branching. It assigns user specified probabilities to every branch, tracks the total branch probability, and truncates branches based on the given pruning/truncation rules to avoid an unmanageable number of scenarios. The function of a dynamic APET developed includes prediction of the conditions, timing, and location of containment failure or bypass leading to the release of radioactive material, and calculation of probabilities of those failures. Thus, scenarios that can potentially lead to early containment failure or bypass, such as through accident induced failure of steam generator tubes, are of particular interest. Also, the work is focused on treatment of uncertainties in severe accident phenomena such as creep rupture of major RCS components, hydrogen burn, containment failure, timing of power recovery, etc. Although the ADAPT methodology (Analysis of Dynamic Accident Progression Trees) could be applied to any severe accident analysis code, in this dissertation the approach is demonstrated by applying it to the MELCOR code [1]. A case study is presented involving station blackout with the loss of auxiliary feedwater system for a pressurized water reactor. The specific plant analyzed is the Zion Nuclear Power Plant, which is a Westinghouse-designed system that has been decommissioned.
Advisors/Committee Members: Aldemir, Tunc.
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21.
Hawn, David Phillip.
Development of a Dynamic Model of a Counterflow Compact Heat Exchanger for Simulation of the GT-MHR Recuperator using MATLAB and Simulink.
Degree: MS, Nuclear Engineering, 2009, Ohio State University
► A computational model was developed to determine the dynamic behavior of counter…
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▼ A computational model was developed to determine the dynamic behavior of counter flow compact heat exchangers. This code was written with the intention of becoming a component of a larger system dynamics model of a Brayton cycle nuclear power plant. Several configurations for the GT-MHR recuperator were analyzed, but the code can easily be modified to analyze many types of compact heat exchangers with a variety of applications. Helium was the working fluid used in this project, but the code can be modified to use other gases. This code was written in Matlab and Simulink but the methods outlined in this report could be easily reapplied in other programming languages. This code is also useful for designing counter flow compact heat exchangers in general.In this model the heat exchanger is discretized in time and in space. The resolution of the discretization is defined by the user. Helium properties are reevaluated for each volume before each time step. The dynamic inputs to the model are the inlet temperature, mass flow rate and pressure for each side of the heat exchanger. This model assumes low Mach number flows and treats the propagation of pressure and mass flow rate changes as instantaneous. The outlet temperature and pressure drop for each side is determined. The results of the simulation were successfully validated against results available in the literature. Contact the author for a copy of this code.
Advisors/Committee Members: Blue, Thomas.
Subjects: Mechanical engineering; Nuclear physics
Keywords: compact heat exchanger; recuperator; system dynamics; GT-MHR; transient analysis
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22.
Hawn, David Phillip.
The Effects of High Temperature and Nuclear Radiation on the Optical Transmission of Silica Optical Fibers.
Degree: PhD, Nuclear Engineering, 2012, Ohio State University
► Distributed measurements made with fiber optic instrumentation have the potential to revolutionize…
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▼ Distributed measurements made with fiber optic instrumentation have the potential to revolutionize data collection for facility monitoring and process control in industrial environments. Dozens of sensors etched into a single optical fiber can be used to instrument equipment and structures so that dozens of spatially distributed temperature measurements, for example, can be made quickly using one optical fiber. Optically based sensors are commercially available to measure temperature, strain, and other physical quantities that can be related to strain, such as pressure and acceleration. Other commercially available technology eliminates the need to etch discrete sensors into an optical fiber and allows temperature measurements to be made along the length of an ordinary silica fiber. Distributed sensing with optical instrumentation is commonly used in the petroleum industry to measure the temperature and pressure profiles in down hole applications. The U.S. Department of Energy is interested in extending the distributed sensing capabilities of optical instrumentation to high temperature reactor radiation environments. For this technology extension to be possible, the survivability of silica optical fibers needed to be determined in this environment. In this work the optical attenuation added to silica optical fiber exposed simultaneously to reactor radiation and temperatures to 1000°C was experimentally determined. Optical transmission measurements were made in-situ from 400nm-2300nm. For easy visualization, all of the results generated in this work were processed into movies that are available publicly [1]. In this investigation, silica optical fibers were shown to survive optically and mechanically in a reactor radiation environment to 1000°C. For the combined high temperature reactor irradiation experiments completed in this investigation, the maximum attenuation increase in the low-OH optical fibers was around 0.5db/m at 1550nm and 0.6dB/m at 1300nm. The radiation induced optical attenuation primarily affected wavelengths less than 1000nm and this attenuation cannot be avoided in silica. Thermal effects dominated the increase in attenuation at wavelengths above 1000nm and it may be possible to mitigate these effects. Fortuitously, commercial optical instrumentation typically utilizes wavelengths centered around 1300nm and 1550nm where the radiation induced attenuation was minimal. The maximum continuous use temperature of silica optical fiber may be limited to 900°C with intermittent use to 1000°C. The silica optical fibers tested in this project are inexpensive and commercially available. Optical sensors were not tested in this project and development and testing of radiation hard optical sensors is recommended as future work.
Advisors/Committee Members: Blue, Thomas.
Subjects: Experiments; Materials Science; Nuclear Engineering; Optics; Radiation
Keywords: silica optical fiber; high temperature silica; radiation induced attenuation; fiber optic instrumentation; in-situ optical attenuation
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23.
Kandlakunta, Praneeth.
A Proof-of-Principle Investigation for a Neutron-Gamma Discrimination Technique in a Semiconductor Neutron Detector.
Degree: MS, Nuclear Engineering, 2012, Ohio State University
► Gadolinium (Gd) is an efficient thermal neutron conversion material due to the…
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▼ Gadolinium (Gd) is an efficient thermal neutron conversion material due to the superior thermal neutron absorption cross-section predominantly composed of 157Gd, which is also 15.65% abundant in natural Gd. A thermal neutron capture by 157Gd results in an excited 158Gd nucleus. The de-excitation of 158Gd* involves the emission of prompt gamma rays with competing internal conversion (IC) electrons. Following the expulsion of conversion electrons from the atomic shells of 158Gd, the excitation energy of the atom is released through the emission of Auger electrons and characteristic x-rays. The low energy conversion electrons and Auger electrons are considered the principal component of neutron-induced signal in a Gd-based thin film semiconductor. Besides possessing a high thermal neutron absorption cross-section, Gd also has a good interaction probability for high/medium energy gamma rays, owing to its high Z (64) number. A Gd atom activated by an external high energy gamma ray leads to the emission of characteristic x-rays that come primarily from the K-shell. These x-rays have a fairly low energy (43.0 keV, 42.3 keV for Kα1, Kα2 respectively) compared to those of the prompt gamma rays that are emitted following a neutron capture. Thin film semiconductors, although transparent to high energy gamma rays, are comparatively sensitive to low energy gamma rays and x-rays. Hence, it is supposed that a thin film semiconductor neutron detector using Gd as a neutron convertor receives greater interference from low energy x-rays that are emitted following gamma ray activation in Gd, than that from high energy background gamma rays. Thus, due to the presence of an inherent gamma ray background, separation of the neutron-induced signal from a gamma/x-ray induced signal is central to a semiconductor neutron detector employing Gd as the conversion material. A method of separation of these two signals by means of a current subtraction technique has been proposed. This gamma ray rejection scheme presents two identical semiconductor detectors separated by a thin Gd foil and a polyethylene thin layer. In the presence of a mixed neutron and gamma ray environment, a subtraction of signals resulting from these two detectors generates a ‘neutron only’ induced signal. Nevertheless, an experimental validation will reinforce the abovementioned supposition and provide substantiation of the same. The objective of this research is thus to validate the principle proposed for gamma ray rejection in a thin film semiconductor neutron detector based on Gd. As the first stage, an experimental setup was designed and constructed to perform the required radiation measurements. In the second phase, preliminary measurements were performed to calibrate the instrumentation system and to gain expertise on using the signal processing electronics. In the final phase, a mixed beta-gamma measurement using two silicon detectors was performed in order to simulate a neutron-gamma discrimination scenario in a Gd based semiconductor detector. The output energy spectra encompassed a mixed beta-gamma spectrum from an unshielded silicon detector and a gamma ray only spectrum from a shielded silicon detector. Subtraction of the two spectra generated a beta-only spectrum representing a detector’s response to the IC and Auger electrons from Gd.
Advisors/Committee Members: Cao, Lei.
Subjects: Nuclear Engineering
Keywords: neutron detection, semiconductor neutron detector, neutron converter, neutron-gamma discrimination
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24.
Kennedy, Ryanne Ariel.
Quantifying Uncertainty in Reactor Flux/Power Distributions.
Degree: PhD, Nuclear Engineering, 2011, Ohio State University
► The design and development of a conceptual system for power measurements in…
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▼ The design and development of a conceptual system for power measurements in a reactor core using in-core sensors has been an ongoing focus of research performed at The Ohio State University. Previous work focused on the development of software that constructs three-dimensional core power distribution using signals from an array of Constant Temperature Power Sensors distributed in the reactor core. The monitoring system processes the sensor signals from the reactor core in such a manner that probabilistic information, in the form of probability distribution functions (pdfs), of reactor power density is obtained at the sensor locations. The desired pdfs of power density at sensor locations were obtained through the implementation of the estimation algorithm Dynamic System Doctor (DSD). Additional work in this area proposed methods to interpolate the pdfs at the measurement points for uniform core compositions. The interpolation methods presented in this thesis extend estimation of the uncertainty in power/flux distribution to heterogeneous cores by combining the knowledge from in-core sensors and reactor core neutron transport calculations. Two significant benefits of such an estimation process are the following: 1. It allows quantification of the uncertainty on peaking factors as well as global power/flux distributions and hence evaluation of potential for power upgrades 2. It provides data for a best-estimate quantification of design margins. The interpolation method is an adaptation of an interpolation procedure presented by L. Read and utilizes the flux/power distributions obtained from a realistic reactor core model as interpolating functions. The pdfs at the interpolated power/flux values are calculated using a weighted linear interpolation of the sensor pdfs. Four weighting schemes are considered for the power/flux interpolation algorithm. The proposed interpolation methods and each weighting scheme are tested using flux data from models of the Ohio State University Research Reactor (OSURR). The objective of the testing of each scheme is to determine its accuracy/conservatism, and to gauge whether the interpolation is physically representative of the process under consideration. Different methods are compared as to their ability to accurately estimate the pdfs at different positions in the reactor core. In view of the uncertainty in the current material composition of OSURR due to burnup of fuel during the past two decades, the Monte Carlo code MCNP5 is used to generate the “experimental data”. This approach also allows control of the uncertainty on the tally results by changing the number of neutron simulations performed. For the test problems, the OSURR core model flux distributions used as distribution interpolating functions (DIF) were generated by PENTRAN (Parallel Environment Neutral particle Transport) 3-D discrete ordinates code (version 9.36k). Three example problems were used to test the linear interpolation schemes. In these three examples the interpolating method is applied to the flux distribution in and around the Central Irradiation Facility (CIF) of the OSURR. A forward flux based interpolation scheme was found to be accurate in estimating the uncertainty in all three example cases. The results indicate that the detector configuration and the fidelity of the reactor core model which supplied the DIF are the key features in determining the accuracy of all four of the interpolation schemes examined.
Advisors/Committee Members: Aldemir, Tunc.
Subjects: Nuclear Engineering
Keywords: nuclear reactor, flux, uncertainty quantification, reactor physics
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25.
Khorsandi, Behrooz.
Modeling of displacement damage in silicon carbide detectors resulting from neutron irradiation.
Degree: PhD, Nuclear Engineering, 2007, Ohio State University
► There is considerable interest in developing a power monitor system for Generation…
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▼ There is considerable interest in developing a power monitor system for Generation IV reactors (for instance GT-MHR). A new type of semiconductor radiation detector is under development based on silicon carbide (SiC) technology for these reactors. SiC has been selected as the semiconductor material due to its superior thermal-electrical-neutronic properties. Compared to Si, SiC is a radiation hard material; however, like Si, the properties of SiC are changed by irradiation by a large fluence of energetic neutrons, as a consequence of displacement damage, and that irradiation decreases the life-time of detectors. Predictions of displacement damage and the concomitant radiation effects are important for deciding where the SiC detectors should be placed. The purpose of this dissertation is to develop computer simulation methods to estimate the number of various defects created in SiC detectors, because of neutron irradiation, and predict at what positions of a reactor, SiC detectors could monitor the neutron flux with high reliability. The simulation modeling includes several well-known – and commercial – codes (MCNP5, TRIM, MARLOWE and VASP), and two kinetic Monte Carlo codes written by the author (MCASIC and DCRSIC). My dissertation will highlight the displacement damage that may happen in SiC detectors located in available positions in the OSURR, GT-MHR and IRIS. As extra modeling output data, the count rates of SiC for the specified locations are calculated. A conclusion of this thesis is SiC detectors that are placed in the thermal neutron region of a graphite moderator-reflector reactor have a chance to survive at least one reactor refueling cycle, while their count rates are acceptably high.
Advisors/Committee Members: BLUE, Thomas E.
Subjects: Engineering, Nuclear
Keywords: Displacement damage, silicon carbide, Monte Carlo methids, count rate, GT-MHR, IRIS
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26.
Kulisek, Jonathan Andrew.
The Effects of Nuclear Radiation on Schottky Power Diodes and Power MOSFETs.
Degree: PhD, Nuclear Engineering, 2010, Ohio State University
► NASA is exploring the potential use of nuclear reactors as power sources…
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▼ NASA is exploring the potential use of nuclear reactors as power sources for future space missions. These missions will require electrical components, consisting of power circuits and semiconductor devices, to be placed in close vicinity to the reactor, in the midst of a high neutron and gamma-ray radiation field. Therefore, the primary goal of this research is to examine the effects of a mixed neutron and gamma-ray radiation field on the static and dynamic electrical performance of power Schottky diodes and power MOSFETs in order to support future design efforts of radiation-hard power semiconductors and circuits. In order to accomplish this goal, commercial Si and 4H-SiC Schottky barrier power diodes were irradiated in the mixed neutron and gamma-ray radiation field of The Ohio State University research reactor (OSURR). The forward I-V characteristics were measured before and immediately after each successive radiation dose and the carrier-removal rates were compared, on the basis of NIEL, to a previous study, for which the same diode models were irradiated with a 203 MeV proton beam. In addition, a number of SiC Schottky barrier diodes were also irradiated in the OSURR and subsequently functionally tested in half-wave rectifier circuits, for which the voltage and current waveforms in the circuit were recorded. The results from the functional testing of these half-wave rectifier circuits were analyzed using results from I-V characterization, PSpice simulations, and an analytical formulation. In addition, boost and buck converters containing commercial power MOSFETs and Schottky diodes that were irradiated to various doses in a mixed neutron and gamma-ray radiation field, were tested. In addition to overall circuit performance in terms of output voltage and efficiency, the individual voltage and current waveforms of the MOSFET and diode in each circuit were examined. Radiation-induced changes in the switching characteristics of the MOSFETs were observed. Furthermore, changes in overall circuit performance and increased power dissipation in the MOSFETs during the over-voltage turn-off transient and on-state conduction portions of the switching cycle were observed. In addition to I-V characterization of the MOSFETs and diodes, PSpice simulations were performed in order to aid in the analysis and interpretation of the experimental results.
Advisors/Committee Members: Blue, Thomas.
Subjects: Electrical engineering; Nuclear physics
Keywords: power MOSFET, SiC, Si, Schottky diode, NIEL, displacement damage, TID, DC-to-DC switching converters, radiation effects
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27.
Lin, Jonathan Lee.
Evaluation of 18F-FDG PET Agent in Cardiac Gated Imaging.
Degree: MS, Nuclear Engineering, 2012, Ohio State University
► Gated cardiac imaging (GCI) is used in oncology and cardiology to improve…
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▼ Gated cardiac imaging (GCI) is used in oncology and cardiology to improve resolution impaired by cardiac motion. Both subjects require specific imaging techniques to accurately portray the inner functionality of the human heart. Positron emission tomography (PET) is a sensitive imaging technique that details the cardiac cycle by recording the three dimensional distribution of radiotracer photon emissions. Unlike the commonly used computed tomography (CT) method which transmits source particles through the target, the PET source emits positrons from within the object, capitalizing on opposing photon emissions for anomaly detection. Within PET imaging, gated cardiac PET using an 18F-fluorodeoxyglucose (18F-FDG) source is a quantitative imaging technique which utilizes a gating technique that reduces signal blurring in the cardiac cycle. This paper evaluates 18F-FDG uptake within gated cardiac PET while developing an elementary interactive data language (IDL) program for gated cardiac PET data analysis. Digital Imaging and Communications in Medicine (DICOM) standards were used to understand PET imaging output, and DICOM conformance was used to determine the DICOM tags required for analysis. The IDL program has two objectives: to read DICOM tags integral to gated PET into the program and to develop a user interface which displays gated PET images in correct sequence. Five gated PET datasets were used to compare and verify the output of the IDL program to the output of verified and authorized EBW/LEO DICOM display workstations. Quantified results show an average end-diastolic volume (EDV) of 25.6 ml, average end-systolic volume (ESV) of 13.8 ml, and a left ventricle ejection fraction (EF) of 51.8 percent. Additionally, results show that the IDL program’s 1-dimensional (1D) trans-axial display matches the output from authorized workstations. In particular, the IDL program provides a graphical user interface (GUI) which facilitates visibility of gated images organized by frame and by slice, allowing the user to easily customize the axial view for further analysis. Thus, the IDL program can be utilized as a promising tool to allow multi-pharmaceutical, multi-modality medical imaging technology users to obtain information of interest utilizing an easy-to-use format.
Advisors/Committee Members: Cao, Lei.
Subjects: Medical Imaging
Keywords: Gated; Cardiac; imaging; GCI; PET; CT; 18F-FDG; IDL; DICOM; ejection; fraction; EF
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28.
Mandelli, Diego.
SCENARIO CLUSTERING AND DYNAMIC PROBABILISTIC RISK ASSESSMENT.
Degree: PhD, Nuclear Engineering, 2011, Ohio State University
► The recent trend to use a best estimate plus uncertainty (BEPU) approach…
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▼ The recent trend to use a best estimate plus uncertainty (BEPU) approach to nuclear reactor safety analysis instead of the traditional conservative approach can produce very large amounts of data. Hence, the need for methodologies able to handle high volumes of data in terms of both cardinality (due to the high number of uncertainties included in the analysis) and dimensionality (due to the complexity of systems) arises. Clustering methodologies offer powerful tools that can help the user to identify groups of scenario that are representative of the original data set and, thus, can reduce the effort involved in data analysis. Scenario clustering aims to: a)~identify the scenarios that have a similar behavior (i.e., identify the most evident classes), b)~decide for each event sequence to which cluster it belongs (i.e., classification), and c)~perform the analysis of each cluster. The main objective of this dissertation is to show how it is possible to accomplish these three objectives by using clustering methodologies to the scenarios generated by safety analysis codes. Several clustering algorithms are developed, evaluated and compared using different types of data sets. Mode-seeking clustering algorithms such as Mean-Shift are proven to be well suited for the scenario analysis. The Mean-Shift algorithm is a kernel-based, non-parametric density estimation technique that is used to find the modes of an unknown distribution, which correspond to regions with highest data density. The obtained cluster centers represent the most representative scenarios from the original data set and the analysis can be now carried out on the smaller set of representative scenarios. The specific types of data under consideration are those generated using the dynamic event tree (DET) approach for nuclear power reactor transients which are described by a large set of state variables (i.e., temperature, pressure of specific nodes in the plant simulator) and information regarding the status of specific components/systems. Several examples are presented in order to illustrate the applications of clustering algorithms to data generated by DET. In addition, pre-processing of the raw data and data reduction techniques are described and compared.
Advisors/Committee Members: Aldemir, Tunc.
Subjects: Nuclear Engineering
Keywords: Scenario analysis, even trees; clustering; dynamic methodology; PRA; DET
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29.
Metzroth, Kyle G.
A Comparison of Dynamic and Classical Event Tree Analysis for Nuclear Power Plant Probabilistic Safety/Risk Assessment.
Degree: PhD, Nuclear Engineering, 2011, Ohio State University
► The development of methods of dynamic probabilistic risk assessment (PRA) is an…
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▼ The development of methods of dynamic probabilistic risk assessment (PRA) is an ongoing topic of research at the Ohio State University. Recently, the ADAPT (Analysis of Dynamic Accident Progression Trees) software tool was developed to provide a flexible framework in which to perform a dynamic event tree analysis. Dynamic PRA methodologies have the advantage over conventional PRA methodologies in that a more realistic and mechanistically consistent analysis can be performed of a system in question. Dynamic PRA methodologies are designed to take the timing of events explicitly into account which can become very important especially when uncertainties in complex phenomena are considered. Despite the advantages that dynamic event methodologies offer, there is still considerable question in the community as to how dynamic methodologies can provide “better” results than classical methods, whether there is really a need the detailed modeling that dynamic methodologies provide within the context of a full PRA, and whether the implementation of dynamic methodologies on a real system is practical as dynamic methodologies can be computationally expensive. The purpose of this work is to address those concerns just noted by performing a comparison of the results obtained for a particular scenario on a real system by using classical PRA analysis and a parallel analysis performed using a particular dynamic PRA method. In late 1980’s the NUREG-1150 study was commissioned to perform a full PRA using the best methods available at the time of five U.S. nuclear power plants. The power plants that were chosen were: Surry Unit 1 (PWR), Zion Unit 1 (PWR), Grand Gulf Unit 1 (BWR), Peach Bottom Unit 1 (BWR), and Sequoyah Unit 1 (PWR). For each of these plants, a detailed analysis of all systems and potential accident pathways was performed using conventional PRA methodology. For this study, the results obtained in NUREG-1150 for the Zion Unit 1 plant will be compared to the results obtained in a dynamic PRA analysis. Specifically, the results obtained for one of the initiating events examined in NUREG-1150, namely the Loss of Offsite Power (LOSP) initiating event with loss of all diesel generators (commonly known as a Station-Blackout (SBO) Event), will be compared. Data from all supporting documentation on the Zion Unit 1 NUREG-1150 analysis have been gathered and a comparable dynamic model will be built and executed using the MELCOR severe accident analysis code. The end-goals of this research are to 1) evaluate the advantages and disadvantages of both conventional and dynamic PRA with respect to one another, 2) compare the numerical results of the conventional and dynamic analysis with respect to the chosen event’s contribution to core damage frequency and large early-release frequency, and 3) to make a comparison of the accident sequences generated by both analyses to determine if dynamic analysis shows additional risk-significant scenarios not discovered by classical methods. Results from the dynamic analysis show consistency with the classical PRA results. However, the dynamic analysis was able to provide additional resolution and detail for some classical plant damage stages. In addition, new accident sequences which were not considered by the classical analysis were discovered. Dynamic event tree analysis proved to be a powerful tool in modeling plant response in a physically consistent manner given an input probabilistic model and provided additional insight into the potential accident progression.
Advisors/Committee Members: Aldemir, Tunc.
Subjects: Nuclear Engineering
Keywords: Probabilistic Risk Assessment; Dynamic Probabilistic Risk Assessment; Nuclear Power Plant Risk Analysis
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30.
Mkhosi, Margaret Msongi.
Computational fluid dynamics analysis of aerosol deposition in pebble beds.
Degree: PhD, Nuclear Engineering, 2007, Ohio State University
► The Pebble Bed Modular Reactor is a high temperature gas cooled reactor…
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▼ The Pebble Bed Modular Reactor is a high temperature gas cooled reactor which uses helium gas as a coolant. The reactor uses spherical graphite pebbles as fuel. The fuel design is inherently resistant to the release of the radioactive material up to high temperatures; therefore, the plant can withstand a broad spectrum of accidents with limited release of radionuclides to the environment. Despite safety features of the concepts, these reactors still contain large inventories of radioactive materials. The transport of most of the radioactive materials in an accident occurs in the form of aerosol particles. In this dissertation, the limits of applicability of existing computational fluid dynamics code FLUENT to the prediction of aerosol transport have been explored. The code was run using the Reynolds Averaged Navier-Stokes turbulence models to determine the effects of different turbulence models on the prediction of aerosol particle deposition. Analyses were performed for up to three unit cells in the orthorhombic configuration. For low flow conditions representing natural circulation driven flow, the laminar flow model was used and the results were compared with existing experimental data for packed beds. The results compares well with experimental data in the low flow regime. For conditions corresponding to normal operating of the reactor, analyses were performed using the standard k-å turbulence model. From the inertial deposition results, a correlation that can be used to estimate the deposition of aerosol particles within pebble beds given inlet flow conditions has been developed. These results were converted into a dimensionless form as a function of a modified Stokes number. Based on results obtained in the laminar regime and for individual pebbles, the correlation developed for the inertial impaction component of deposition is believed to be credible. The form of the correlation developed also allows these results to be applied to pebble beds of different porosities. The effect of turbulence on the deposition of aerosols was analyzed using the discrete random walk model. The results obtained with k-å turbulence model show high deposition of aerosols at low particle diameters. To validate the results in this regime, detailed experimental work is needed.
Advisors/Committee Members: Denning, Richard S.
Subjects: Engineering, Nuclear
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