PhD, University of Cincinnati, 2001, Engineering : Nuclear and Radiological Engineering

In this research a Stochastic Parameter Simulation System (SPSS) computer program employing wavelet techniques was developed. The SPSS was designed to fulfill two key functional requirements:1. To be able to analyze any steady state plant signal, decompose it into its deterministic and stochastic components, and then reconstruct a new, simulated signal that possesses exactly the same statistical noise characteristics as the actual signal; and 2. To be able to filter out the principal serially-correlated, deterministic components from the analyzed signal so that the remaining stochastic signal can be analyzed with signal validation tools that are designed for signals drawn from independent random distributions. The results obtained using SPSS were compared to those obtained using the Argonne National Laboratory Reactor Parameter Simulation System (RPSS) which uses a Fourier transform methodology to achieve the same objectives. RPSS and SPSS results were compared for three sets of stationary signals, representing sensor readings independently recorded at three nuclear power plants. For all of the recorded signals, the wavelet technique provided a better approximation of the original signal than the Fourier procedure. For each signal, many wavelet-based decompositions were found by the SPSS methodology, all of which produced white and normally distributed signal residuals. In most cases, the Fourier-based analysis failed to completely eliminate the original signal serial-correlation from the residuals. The reconstructed signals produced by SPSS are also statistically closer to the original signal than the RPSS reconstructed signal. Another phase of the research demonstrated that SPSS could be used to enhance the reliability of the Multivariate Sensor Estimation Technique (MSET). MSET uses the Sequential Probability Ratio Test (SPRT) for its fault detection algorithm. By eliminating the MSET residual serial-correlation in the MSET training phase, the SPRT user-defined f (open full item for complete abstract)

Committee: Dr. John Christenson (Advisor) Subjects: Engineering, Nuclear

PhD, University of Cincinnati, 2001, Engineering : Nuclear and Radiological Engineering

The Monte Carlo code MCNP4B was used to investigate the radiation dose distribution in several areas of intravascular brachytherapy that would be very difficult or impossible to investigate experimentally or analytically. A model for a liquid-filled balloon catheter was developed and validated by comparing the results of Monte Carlo simulations with experimental measurements made in a tissue equivalent phantom. The dose distribution in the coronary vessel wall from a liquid-filled balloon catheter containing the following radioisotopes was examined: Y-90, Re-188, P-32, Re-186, Sm-153, In-111, and Tc-99m. At 0.5 mm from the vessel surface, the beta emitters deliver a higher dose per unit cumulated activity than the gamma emitters. The gamma emitters, however, deliver a dose that is more uniform throughout the vessel wall. Eight sizes of balloon catheters were modeled and the effect of balloon catheter size on the radiation dose distribution was examined. Effects that perturb the radiation dose distribution for a liquid-filled balloon catheter were investigated. These perturbing effects were: (1) an air bubble within the balloon catheter, (2) contrast media within the balloon catheter, (3) a Palmaz-Schatz stent surrounding the balloon catheter, and (4) the deflation of the balloon catheter. The radiation dose distribution produced from a P-32 coated Palmaz-Schatz stent was investigated by developing a model with P-32 deposited on the stent surface to a depth of one micron, and calculating the dose delivered to the coronary vessel wall. High dose fluctuations were observed near the surface of the stent, but these fluctuations leveled off at depth in the vessel wall. The radiation dose distribution for the case of direct injection of radioisotopes into the coronary vessel wall using the Infiltrator angioplasty balloon catheter (IABC) was also investigated. The source distribution produced by the IABC was modeled for two configurations within the vessel wall: (1) uniform (open full item for complete abstract)

Committee: Dr. Henry Spitz (Advisor) Subjects: Engineering, Nuclear

PhD, University of Cincinnati, 2005, Engineering : Nuclear and Radiological Engineering

The ability to accurately predict the multiplication factor (k eff ) of a nuclear reactor core as a function of exposure continues to be an elusive task for core designers despite decades of advances in computational methods. The difference between a predicted eigenvalue (target) and the actual eigenvalue at critical reactor conditions is herein referred to as the “eigenvalue drift.” This dissertation studies exposure-dependent eigenvalue drift using MCNP-based fuel management analysis of the ORNL High Flux Isotope Reactor core. Spatial-dependent burnup is evaluated using the MONTEBURNS and ALEPH codes to link MCNP to ORIGEN to help analyze the behavior of k eff as a function of fuel exposure. Understanding the exposure-dependent eigenvalue drift of a nuclear reactor is of particular relevance when trying to predict the impact of major design changes upon fuel cycle behavior and length. In this research, the design of an advanced HFIR core with a fuel loading of 12 kg of 235 U is contrasted against the current loading of 9.4 kg. The goal of applying exposure dependent eigenvalue characterization is to produce a more accurate prediction of the fuel cycle length than prior analysis techniques, and to improve our understanding of the reactivity behavior of the core throughout the cycle. This investigation predicted a fuel cycle length of 40 days, representing a 50% increase in the cycle length in response to a 25% increase in fuel loading. The average burnup increased by about 48 MWd/kg U and it was confirmed that the excess reactivity can be controlled with the present design and arrangement of control elements throughout the core's life. Another major design change studied was the effect of installing an internal beryllium reflector upon cycle length. Exposure dependent eigenvalue predictions indicate that the actual benefit could be twice as large as that originally assessed via beginning-of-life (BOL) analyses.

Committee: G. Ivan Maldonado (Advisor) Subjects: Engineering, Nuclear

PhD, University of Cincinnati, 2005, Engineering : Nuclear and Radiological Engineering

The Boltzmann equation, which is often used to describe neutral particle transport process, is impossible to solve analytically in most cases. Solving the equation by a numerically method could also be very time-consuming for some problems. Therefore, various approximation methods have been developed for particle transport calculations. In this dissertation, a new Flux-Limited diffusion method (DM) is investigated as a low-order approximation to the transport equation. Specifically, the new Flux-Limited diffusion method (DM) is derived based on the Minerbo maximum entropy Eddington factor and corresponding boundary conditions are developed. Performance of the DM method is tested for a variety of particle transport problems: including steady-state problems, transient problems, one-dimensional problems, two-dimensional problems, flux distribution problems, eigen-value problems, one-group problems, and two-group problems. The tests demonstrate that compared with other approximations, the DM method has the following good features. (1) The DM method is a low order approximation as the classic diffusion method but can improve the accuracy of classic diffusion method by a factor up to 10. (2) The DM method has relatively fast computing speed compared with other high order approximations and its computation cost is not too expensive compared with the classic diffusion method. (3) The DM method can predict proper traveling speed of particles in time-dependent problems. (4) The DM method is valid for neutral particle transport in systems containing strong absorbers, regions near particle sources or boundaries, and systems where strongly anisotropic scattering of particles occur. (5) The DM method is superior to other Flux-Limited diffusion methods available in the literature. (6) The solutions of the DM method are numerically stable in all the circumstances. Overall, the DM method is promising as a low-order approximation for particle transport calculations in some applicatio (open full item for complete abstract)

Committee: Bingjing Dr. Su (Advisor) Subjects: Engineering, Nuclear

PhD, University of Cincinnati, 2004, Engineering : Nuclear and Radiological Engineering

The Pebble Bed Reactor (PBR) is a helium-cooled, graphite-moderated high temperature nuclear power reactor. In addition to its inherently safe design, a unique feature of this reactor is its multipass fuel cycle in which graphite fuel pebbles (of varying enrichment) are randomly loaded and continuously circulated through the core until they reach their prescribed end-of-life burnup limit (~ 80,000-100,000 MWD/MTU). Unlike the situation with conventional light water reactors (LWRs), depending solely on computational methods to perform in-core fuel management will be highly inaccurate. As a result, an on-line measurement approach becomes the only accurate method to assess whether a particular pebble has reached its end-of-life burnup limit. In this work, an investigation was performed to assess the feasibility of passive gamma-ray spectrometry assay as an approach for on-line interrogation of PBR fuel for the simultaneous determination of burnup and enrichment on a pebble-by-pebble basis. Due to the unavailability of irradiated or fresh pebbles, Monte Carlo simulations were used to study the gamma-ray spectra of the PBR fuel at various levels of burnup. A pebble depletion calculation was performed using the ORIGEN code, which yielded the gamma-ray source term that was introduced into the input of an MCNP simulation. The MCNP simulation assumed the use of a high-purity coaxial germanium detector. Due to the lack of one-group high temperature reactor cross sections for ORIGEN, a heterogeneous MCNP model was developed to describe a typical PBR core. Subsequently, the code MONTEBURNS was used to couple the MCNP model and ORIGEN. This approach allowed the development of the burnup-dependent, one-group spectral-averaged PBR cross sections to be used in the ORIGEN pebble depletion calculation. Based on the above studies, a relative approach for performing the measurements was established. The approach is based on using the relative activities of Np-239/I-132 in combination w (open full item for complete abstract)

Committee: Ayman Hawari (Advisor) Subjects: Engineering, Nuclear

PhD, University of Cincinnati, 2002, Engineering : Nuclear and Radiological Engineering

On-line signal validation is essential for safe and economic operations of a complicated industrial system such as a nuclear power plant. Various signal validation methods based on empirical signal estimation have been developed and successfully used. The first part of the thesis addresses a common and unavoidable problem for these methods -fault propagation, which causes false identification of healthy signals as faulty ones because of the faults existing in other signals. This effect is especially serious when faults occur in multiple signals and/or during system transient. A sub-grouping technique is presented in the thesis to prevent the effect of fault propagation in general signal validation methods. Specifically, two methods, Subgroups Consistency Check (SCC) and Subgroups Voting (SV), are developed. Their effectiveness is demonstrated by using a well-known Multivariate State Estimation technique (MSET) as a general method of signal estimation. To further improve the performance of MSET estimation, a procedure called Feedback Once (FBO) is also developed. All these new methods are tested and compared with MSET by using real transient data from a reactor startup process in a nuclear power plant. The results show that false identification of signals caused by fault propagation is significantly reduced by the two sub-grouping methods and the FBO method is able to improve the performance of MSET estimation to some extent. The results demonstrate that implementation of these new methods can lead to an improved signal validation technique that remains effective even when faults occur in multiple signals during system transients. The other major contribution is on the improvement of statistical test used for signal validation. Sequential Probability Ratio Test (SPRT) is a popular method that has been widely used in many signal validation methods. However, the assumption of SPRT is too stringent to satisfy in practice, which may cause the false identification rate ex (open full item for complete abstract)

Committee: Dr. Bingjing Su (Advisor) Subjects: Engineering, Nuclear

Doctor of Philosophy, The Ohio State University, 2007, Nuclear Engineering

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-a 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 por (open full item for complete abstract)

Committee: Richard Denning (Advisor) Subjects: Engineering, Nuclear

Doctor of Philosophy, The Ohio State University, 2007, Nuclear Engineering

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.

Committee: Thomas BLUE (Advisor) Subjects: Engineering, Nuclear

Doctor of Philosophy, The Ohio State University, 2006, Nuclear Engineering

In this dissertation, a fast neutron flux monitoring channel, which is based on the use of SiC semiconductor detectors is designed, modeled and experimentally evaluated as a power monitor for the Gas Turbine Modular Helium Reactors. A detailed mathematical model of the SiC diode detector and the electronic processing channel is developed using TRIM, MATLAB and PSpice simulation codes. The flux monitoring channel is tested at the OSU Research Reactor. The response of the SiC neutron-monitoring channel to neutrons is in close agreement to simulation results. Linearity of the channel response to thermal and fast neutron fluxes, pulse height spectrum of the channel, energy calibration of the channel and the detector degradation in a fast neutron flux are presented. Along with the model of the neutron monitoring channel, a Simulink model of the GT-MHR core has been developed to evaluate the power monitoring requirements for the GT-MHR that are most demanding for the SiC diode power monitoring system. The Simulink model is validated against a RELAP5 model of the GT-MHR. This dyanamic model is used to simulate reactor transients at the full power and at the start up, in order to identify the response time requirements of the GT-MHR. Based on the response time requirements that have been identified by the Simulink model and properties of the monitoring channel, several locations in the central reflector and the reactor cavity are identified to place the detector. The detector lifetime and dynamic range of the monitoring channel at the detector locations are calculated. The channel dynamic range in the GT-MHR central reflector covers four decades of the reactor power. However, the detector does not survive for a reactor refueling cycle in the central reflector. In the reactor cavity, the detector operates sufficiently long; however, the dynamic range of the channel is smaller than the dynamic range of the channel in the central reflector.

Committee: Thomas Blue (Advisor) Subjects: Engineering, Nuclear