Browsing by Author "Mohamed A. Bourham, Committee Member"

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Advanced Feedwater Control for Next Generation Nuclear Power Systems
(2006-07-06) Shen, Hengliang; Mohamed A. Bourham, Committee Member; Mo-Yuen, Chow, Committee Member; Man-Sung Yim, Committee Member; J. Michael Doster, Committee Chair
In current generation Pressurized Water Reactors (PWRs), the control of Steam Generator level experiences challenges over the full range of plant operating conditions. These challenges can be particularly troublesome in the low power range where the feedwater is highly subcooled and minor changes in the feed flow may cause oscillations in the SG level, potentially leading to reactor trip. Substantial attention has been given to feedwater control systems with recognition of the difficulty of the full range feedwater control problem due to steam generator level shrink-swell phenomena, changes in valve and flow path characteristics, and other nonlinear phenomena over the full range of operating conditions[1]. The IRIS reactor concept adds additional challenges to the feedwater control problem as a result of a steam generator design where neither level or steam generator mass inventory can be measured directly[2]. Neural networks have demonstrated capabilities to capture a wide range of dynamic signal transformation and non-linear problems[3-5]. In this project a detailed engineering simulation of plant response is used to develop and test neural control methods for the IRIS full range feedwater control problem. The established neural feed controller has demonstrated the capability to improve the performance of SG level or mass control under transient conditions and over a wide range of reactor power including abnormal conditions.
Applications of Atmospheric Plasmas
(2009-11-08) Oldham, Christopher John; Mohamed A. Bourham, Committee Member; Jerome J. Cuomo, Committee Chair; J. Michael Rigsbee, Committee Member; Mark A.L. Johnson, Committee Member
Surface modification techniques using plasmas have generally been completed in a low pressure environment due to Pd (pressure x gap distance) considerations influencing the behavior of plasma generation. Generally, plasmas produced in a low pressure environment are of a non-thermal or cold nature. The basic feature of non-thermal plasmas is the majority of electrical energy used to generate the plasma is primarily used to produce energetic electrons for generating chemical species. Low pressure plasmas serve many purposes for materials processing. Since the plasma environment is contained within a closed vessel, the plasma can be controlled very easily. Low pressure plasmas have been used in many industries but the complexity associated with the large pumping stations and limitation to batch processing has motivated new work in the area of atmospheric plasmas. Atmospheric plasmas offer both economic and technical justification for use over low pressure plasmas. Since atmospheric plasmas can be operated at ambient conditions, lower costs associated with continuous processing and a decrease in the complexity of equipment validate atmospheric plasma processing as a next generation plasma-aided manufacturing processes. In an effort to advance acceptance of atmospheric plasma processing into industry, a process was developed, the dielectric barrier discharge (DBD), in order to generate a homogeneous and non-thermal plasma discharge at ambient conditions. The discharge was applied to the reduction of known food borne pathogens, deposition of thin film materials, and modification of lignocellulosic biomass.
Benchmarking Thermal Neutron Scattering in Graphite
(2007-12-21) Zhou, Tong; Wesley E. Snyder, Committee Member; Mohamed A. Bourham, Committee Member; Bernard W. Wehring, Committee Member; Ayman I. Hawari, Committee Chair
The Very High Temperature Reactor (VHTR), one of the Generation IV reactor concepts, is a helium-cooled, graphite-moderated nuclear reactor with a core temperature reaching 1000°C. It can provide high quality process heat for hydrogen production beside power generation and will become deployable around 2030. At such temperature, graphite is an appropriate neutron moderator material due to its high sublimation temperature and high temperature strength. Furthermore, graphite has a large heat capacity and stable structure due to its large thermal inertia. However, the current thermal neutron cross-section libraries of graphite are based on models and data developed in the 1950s and 1960s. Significant discrepancies between measurements and the computational predictions of these libraries were observed. As a result, a study was performed in this dissertation to benchmark modern and traditional thermal neutron scattering libraries of graphite. In this work, a Slowing-Down-Time experiment was designed and performed at the Oak Ridge National Laboratory (ORNL) by using the Oak Ridge Electron Linear Accelerator (ORELA) as a neutron source to study the neutron thermalization in graphite at room and higher temperatures. The MCNP5 code was utilized to simulate the detector responses and help optimize the experimental design including the size of the graphite assembly, furnace, shielding system and detector position. To facilitate such calculation, MCNP5 version 1.30 was modified to enable perturbation calculation using point detector type tallies. By using the modified MCNP5 code, the sensitivity of the experimental models to the graphite total thermal neutron cross-sections was studied to optimize the design of the experiment. Measurements of slowing-down-time spectrum in graphite were performed at room temperature for a 70x70x70 cm graphite pile by using a Li-6 scintillator and a U-235 fission counter at different locations. The measurements were directly compared to the Monte Carlo simulations that use different graphite thermal neutron scattering cross-section libraries. Simulations based on the ENDF/B-VI graphite library were found to have a 30%-40% disagreement with the measurements. In addition to the graphite SDT experiment, which provided the data in the energy region above the graphite Bragg-cutoff energy, transmission experiments were performed for different types of graphite samples using the NIST 8.9 Å beam (located at NG-6) to investigating the energy region below the Bragg-cutoff energy. Measurements confirmed that reactor grade graphite, which is a two phase material (crystalline graphite and amorphous carbon), has different thermal neutron scattering cross section from pyrolytic graphite (crystalline graphite). The experiments presented in this work compliment each other and provide an experimental data set which can be used to benchmark graphite thermal neutron scattering cross section libraries that are generated using different methodologies. Further investigation is necessary.
Creep-Rupture Study of Annealed Zircaloy 4: Stress and Temperature Effects
(2005-11-22) Marple, Brian Wesley; Gerd Duscher, Committee Member; Mohamed A. Bourham, Committee Member; K. Linga Murty, Committee Chair
Zircaloys are widely used as fuel rod cladding in light water reactors (LWRs) because of their low cross-section for absorption of thermal neutrons. Currently, the United States does not permit reprocessing of spent fuel so the primary barrier for the spent fuel in a repository will be the fuel rod cladding. Due to the decay heat of the spent fuel, creep rupture is considered to be the primary cause of failure in spent fuel cladding over the long period of time that it will be stored. A fundamental understanding of the creep mechanisms in Zircaloys is crucial to accurately predicting the integrity of the fuel cladding over long periods of time. Zirconium has a hexagonally close-packed crystal structure and because of this, exhibits creep anisotropy that is affected not only by the texture, but also by temperature, stress, and loading. Since the stress imposed on the spent fuel during long-term storage will be relatively low compared to service conditions, the low stress creep behavior must be characterized and mechanistically understood to avoid non-conservative estimates based on in-pile creep data. In addition, loading of spent fuel in a repository will be due to the internal pressure generated by fission product gasses and from the inert gas introduced at the time of fuel fabrication. This work focuses on the creep rupture behavior and microstructural characterization of annealed Zircaloy-4 at temperatures ranging from 250°C-600°C and stresses from 27 MPa-350 MPa. Typically, fuel assemblies that have been fabricated from Zircaloy-4 are not in the annealed condition. Instead, they are cold-worked and stress relieved (CWSR). Since low stress creep rupture testing would take years at low temperatures, high temperatures are used to observe the effects of low stress in a reasonable amount of time. At such high temperatures, the grain structure of the CWSR material would change drastically. Therefore the material was annealed prior to testing to avoid this complication. Testing on unirradiated material will yield higher strain rates because of irradiation hardening. Therefore, estimates based on unirradiated creep rupture data would be conservative. Prior to testing, optical metallographs were taken to characterize the grain structure. A limited texture study was performed to evaluate the texture coefficients for each direction in the rod. Transmission electron microscopy (TEM) was also performed to characterize the initial dislocation microstructure. After testing, diametric measurements were taken and the strain rate determined. From data at various stresses, the activation energy was derived along with equations predictive of rupture such as the Larson-Miller parameter and the Monkman-Grant relationship. Specimens of interest were selected for optical metallography and to obtain TEM micrographs of the dislocation microstructure. The activation energy deduced was in excellent agreement of that for self-diffusion. Optical metallography showed slight grain elongation in samples tested at high stresses while grains remained equiaxed at low stresses. TEM showed significant sub-grain formation at low stresses and random dislocation organization at higher stresses.
Crystallographic Texture and Creep Anisotropy in Cold Worked and Recrystallized Zirlo
(2005-08-09) Yan, Jinyuan; Ron O.Scattergood, Committee Member; Mohamed A. Bourham, Committee Member; Man-Sung Yim, Committee Member; K. Linga Murty, Committee Chair
Zirlo, a special zircaloy material alloyed with niobium, tin and iron is a successor of Zircaloy-4. Zirlo is materials used in fuel rod cladding, structural and flow mixing grids, instrumentation tubes, and guide thimbles. It increases margin to fuel rod corrosion limits and enhance fuel assembly structural stability in Pressurized Water Reactor. Zirconium and its alloys, being hexagonally close packed, have limited number of slip systems, and exhibit preferred orientations following thermo-mechanical treatments, which result in anisotropic mechanical properties. The objective of this project is to investigate the anisotropic mechanical properties, crystallographic texture, and microstructure of crept zirlo materials. The anisotropic mechanical properties were investigated using uniaxial and biaxial creep tests. The specimen was loaded axially by a dead weight pan, and the hoop stresses was achieved by internally pressurizing the specimen with inert argon. Different axial and hoop stress, which produced different stress ratios (0, 0.67,0.75, 1, and 2) are selected for creep tests at 450°C. The axial displacement was measured by a linear variable differential transducer and the diameter change by a laser extensometer. Creep data are used to determine strain rate ratios vs stress ratios, the anisotropic parameters ( R and P), and creep loci for cold-worked and recrystallized zirlo. The crystallographic textures were characterized in terms of inverse and direct pole figures using X-ray diffraction techniques. Inverse pole figures were constructed for specimens in the rolling direction, transverse direction, and normal direction for both cold worked and recrystallized tubes. Direct pole figures were constructed for specific reflection planes, such as basal (0002), prismatic (10 0) and pyramidal (10 2). Crystallite orientation distribution function (CODF) was derived from the pole figure data. Euler plots were obtained from crystallite orientation distribution coefficients (wlmn ) and subsequently therefore, ideal orientations were calculated. These CODFs were combined with the Lower-Bound model to predict creep anisotropy assuming the dominance of prismatic, basal and pyramidal slip systems. Creep strain rate ratios vs stress ratios, creep loci and anisotropy parameters (R and P) were predicted. The predictions based on the prismatic dominance matche with the experimental data very well. Microstructure of the crept specimens was characterized by Transmission Electron Microscopy for different stress ratios ( 0, 0.75 and 1). The results show mainly dislocations in the matrix with no subgrain formation. The samples tested under equibiaxial loading revealed deformation twins. More detailed work is called for in characterizing the influence of stress-states and stress levels as well as cold work on deformation microstructures.
A Dynamically Polarized Deuteron Target
(2007-12-07) Poole, John Owen; Albert R. Young, Committee Member; Paul R. Huffman, Committee Member; Christopher R. Gould, Committee Chair; Mohamed A. Bourham, Committee Member
A dynamically polarized deuteron target was constructed at the Triangle Universities Nuclear Laboratory for low energy polarized neutron transmission experiments. Theoretical calculations suggested a measurement of the longitudinal spin-dependent total $vec[n]-vec[d]$ cross section difference, $Deltasigma_L$, would provide evidence of three nucleon force effects. The target was operated at a temperature of 0.5K with a recirculating $ˆ3$He evaporation refrigerator and a 2.5T split-pole superconducting magnet which can be mechanically mounted to produce either a longitudinally or tangentially polarized target relative to the beam momentum. The target material consisted of a cube of volume 2.7 ml of either partially (D6) or fully (D8) deuterated 1,2-propanediol chemically doped with EHBA-Cr$ˆV$ to provide the paramagnetic centers for dynamic nuclear polarization with microwaves in the region 69GHz. Polarization was monitored during the experiment using a continuous wave NMR Q-meter capable of both phase sensitive and magnitude detection. A PC running LabVIEW controlled data acquisition, frequency sweep of the digital frequency synthesizer and a novel background cancellation technique. A pre-recorded background signal was subtracted on a channel by channel basis before the NMR spectrum was sampled. Polarization was extracted from the NMR spectra by a fit to a theoretical lineshape. Effects including dispersion were considered. The equal spin temperature (EST) hypothesis was used to determine deuteron polarization indirectly though the measurement of polarization of the residual protons in the partially deuterated sample. Efforts to improve the system stability and signal-to-noise are discussed along with numerical methods, fitting strategy and evaluation of uncertainties.
Growth Optimization and Characterization of Reactively Sputtered Zirconium Nitride Thin Films for III-V Buffer Layer Applications
(2002-11-07) McGregor, David Ross Jr.; Jerome J. Cuomo, Committee Chair; Mohamed A. Bourham, Committee Member; Dennis M. Maher, Committee Member
Zirconium nitride (ZrN) thin films were deposited by reactive dc magnetron sputtering to assess the effects of processing conditions upon film properties. Processing conditions and parameters were optimized to generate films of completely oriented (111) ZrN on silicon to be used as buffer layers for the growth of gallium nitride A single and double Langmuir probe were used to determine trends in electron temperature, ion density, ionization fraction, and floating potential during reactive sputtering of zirconium in argon and nitrogen. Reactive gas concentration, deposition pressure, deposition temperature, cathode current, film thickness and substrate orientation were investigated as variable processing conditions. Four-point probe, scanning electron microscopy (SEM), transmission electron microscopy (TEM), Raman spectroscopy, and x-ray diffraction (XRD) were used to characterize thin films produced. The optimum growth conditions for the (111) oriented growth of ZrN, for this work, were found to occur during reactive magnetron sputtering at a deposition temperature of 500°C, a constant cathode current of 0.5 ampere, a deposition pressure of 15 mTorr, a reactive nitrogen gas concentration of 4% in argon, deposited on (111) oriented silicon, with a thickness on the order of 600 nanometers. Gallium nitride was then deposited on films of ZrN to assess the crystallinity of films produced. The lattice mismatch between (111) oriented ZrN and c-axis oriented GaN was calculated at 1.6%. Microscopic evaluation showed the films to be of columnar structure with dense grains and smooth surfaces. A change in preferred orientation was noticed as a function of increasing film thickness and cathode current and was determined to be due to an increase in ion channeling and bombardment energy.
Investigation of Failed TRISO Fuel Assay Using Gamma-Ray Spectrometry
(2008-04-06) Harp, Jason Michael; Ayman I. Hawari, Committee Chair; Mohamed A. Bourham, Committee Member; Kimbersly S. Weems, Committee Member
TRISO microsphere fuel is the fundamental fuel unit for Very High Temperature Reactors (VHTR). A single TRISO particle consists of an inner kernel of Uranium Oxycarbide surrounded by layers of pyrolytic carbon and silicon carbide. The silicon carbide serves as the primary barrier to the release of fission products into the core. If the silicon carbide layer fails, fission gas, especially Kr and Xe, will begin to escape the failed particle. In order to understand the behavior of TRISO fuel under in-core conditions, a series of experiments is being conducted by Idaho National Lab at the Advanced Test Reactor. AGR-1 is the first of these experiments. It will measure fission product release due to failed TRISO particles. Simulations of this experiment have been conducted at North Carolina State University to develop a method for the analysis of the results of the experiment. The ATR core was simulated using the Monte Carlo code MCNP to calculate the expected neutron energy spectrum for the AGR-1 experimental test train. This spectrum was used to create one-group cross sections for implementation in ORIGEN calculations of the amount of activity produced in the experiment. Several theoretical models have been developed to describe the phenomenon of gas release. While each model is based on similar physics, different models contain unique features that distinguish them from one another. These Release to Birth (R⁄B) models are developed and applied to the activity found in the ORIGEN calculations to create expected release activities. The release activity is used to create gamma-ray spectra that are representative of the different R⁄B models. Expected R⁄B due to a model can be calculated for comparison to the experiment with knowledge of the number of failed particles in the spectra. The comparison of measured to predicted R⁄B ratios gives insight into the physics of release and also helps validate specific models. Direct comparison is possible, but many of the uncertainties associated with direct comparison are nullified through the use of relative indicators. Each R⁄B model has a unique set of indicators that reflect the physical processes simulated in the model. Trends in the model indicators can be matched up with trends in indicators derived from the release spectra to validate either an entire model or validate the need to consider certain parameters in the creation of a complete and successful release to birth model. Gamma spectrometry is a useful tool for the understanding of fission gas release from failed TRISO particles. A better understanding of the processes that influence fission gas release will influence the fuel manufacturing and quality assurance protocols during the continued development of the VHTR. Future work in this area includes experiment in which the conditions can be better controlled to document the effects of temperature and fission rate in the fuel.
Investigation of Prompt Gamma-Ray Neutron Activation Analysis for Determining the Phase Amounts in Multiphase Flow
(2008-06-05) Mutiso, Athanas M; Sharon Lubkin, Committee Member; Robin P. Gardner, Committee Chair; Mohamed A. Bourham, Committee Member
Neutron Capture Measurements on 157Gd and 89Y at DANCE
(2010-03-04) Chyzh, Andrii; John H. Kelley, Committee Member; Mohamed A. Bourham, Committee Member; Christopher R. Gould, Committee Member; Gary E. MItchell, Committee Chair
Neutron capture reactions are of crucial importance for different applications in nuclear physics and nuclear engineering. Common challenges in measuring these reactions are the separation of the $(n,gamma)$ channel from other reaction channels, such as $(n,el)$ and $(n,n')$, that contribute large backgrounds and obscure the $gamma$-ray cascade following neutron capture. Several major facilities focus on neutron capture. Perhaps the best detector for neutron capture measurements is the DANCE (Detector for Advanced Neutron Capture Experiments) array, a $4pi$ array of 160 BaF$_2$ crystals. In the present research two nuclei were investigated with DANCE: $^{157}$Gd and $^{89}$Y. The $^{157}$Gd experiment was performed in 2006. This isotope is known for an enormous $(n,gamma)$ cross section -- the largest one in nature for stable isotopes. $^{157}$Gd has several practical applications, including its use as a shutdown system in nuclear reactors, medical therapy, neutron shielding, etc. There are 7 stable isotopes of Gd; this permits the study of the systematics of phenomena such as the scissors mode resonance as a function of mass and deformation. DANCE also enables the improvement of the resonance spectroscopy, including the level density, and thus the strength functions, essential for calculating neutron reaction rates. $^{89}$Y was measured with DANCE in 2008, and additional data for improved statistics were taken in the following year. The main motivation was to improve the neutron capture cross section. Yttrium is important for stewardship science and its isotopes are used as radchem detectors to infer the neutron flux. The $^{89}$Y$(n,gamma)$ cross section is poorly known, with low accuracy and a limited neutron energy range. The major difficulty is that the elastic cross section is so large compared to neutron capture. This thesis is divided into 3 parts. The 1$^{st}$ part contains a detailed description of the hardware, from the proton linac to DANCE, which is used in both experiments. The data acquisition system collects and performs an on-line analysis of the data on the event-by-event basis. The 2$^{nd}$ part is dedicated to the $^{157}$Gd experiment: how the data were taken, the challenges of the off-line analysis, and comparison of the DANCE results with other existing data, and with statistical model calculations. The 3$^{rd}$ part is about the $^{89}$Y experiment. This covers many of the same experimental topics as those for $^{157}$Gd. In addition this section explains how the neutron scattering background was simulated with other nuclei, and how the absolute neutron flux was determined with the aid of a resonance in $^{197}$Au.
Neutron Capture Reactions on Gadolinium Isotopes
(2010-02-02) Baramsai, Bayarbadrakh; Gary E. Mitchell, Committee Chair; Undraa Agvaanluvsan, Committee Co-Chair; Mohamed A. Bourham, Committee Member; John H. Kelley, Committee Member; Christoper R. Gould, Committee Member
The neutron capture reaction on 155Gd, 156Gd and 158Gd isotopes has been studied with the DANCE calorimeter at Los Alamos Neutron Science Center. The highly segmented calorimeter provided detailed multiplicity distributions of the capture gamma-rays. With this information the spins of the neutron capture resonances have been determined. The new technique based on the statistical pattern recognition method allowed the determination of almost all spins of 155Gd s-wave resonances. The generalized method was tested for s- and p-wave resonances in 94Mo and 95Mo isotopes. The results were compared with previous resonance data as well as results from other methods. The 155Gd(n,gamma)156Gd cross section has been measured for the incident neutrons energy range from 1 eV to 10 keV. The results are in good agreement with other experiments. Neutron resonances parameters were obtained using the multilevel R-matrix code SAMMY. The fitted radiation and neutron widths were compared with the nuclear data library ENDF/B-VII.0 and with a recent experiment at RPI. With the new spin assignments and resonance parameters, level spacings and neutron strength functions were determined for s-wave resonances in 155Gd. The Monte Carlo code DICEBOX was used to simulate the gamma-decay of the compound nuclei 156Gd, 157Gd and 159Gd. The DANCE detector response was taken into account with a GEANT4 simulation. The simulated and experimental spectra were compared to determine suitable model parameters for the photon strength functions (PSFs) and the level density (LD). The shape of the photon strength function which gave the best agreement with the DANCE spectra contained four low-lying Lorentzian resonances, two for the scissors mode and two for the M1 spin flip mode.
Plasma Aided Finishing of Textile Materials
(2005-07-08) Matthews, Suzanne Rodden; Marian G. McCord, Committee Chair; Mohamed A. Bourham, Committee Member; Peter J. Hauser, Committee Member; Samuel Hudson, Committee Member
Surface modification of textile materials extends over a wide range of alterations to provide desired single or multi-features for various applications. It is a highly focused area of research in which alterations to physical and/or chemical properties lead to new textile products that provide new applications or satisfy specific needs. These processes, however, can involve numerous chemicals, some of which are toxic to humans and hazardous to the environment. In an effort to eliminate these harmful chemicals and waste products, surface modification and finishing via plasma treatment has become an attractive alternative, and is the focus of this work. Through analyzing and understanding plasma-substrate interactions, new and novel finishing applications have been developed. These processes include plasma-aided desizing of polyvinyl alcohol, and plasma-aided grafting of antimicrobial agents onto polypropylene nonwoven fabrics. Plasma treatment of PVA films has shown a significant amount of size removal through sputtering mechanisms, as well as increased solubility via chain scission, which further aids in ease of removal. Plasma treatment of PP fabrics has shown a viable pretreatment for free radical grafting of antimicrobial agents without the use of chemical etchants. In addition to new processing methods, this work has also provided an investigation into the development of a generalized solubility model for plasma exposed materials.
Proton Radiative Capture on 48Ti
(2003-12-04) McDevitt, Daniel Bruno; Dr. G.E. Mitchell, Committee Co-Chair; John F. Shriner, Jr., Committee Co-Chair; D. Ronald Tilley, Committee Member; Mohamed A. Bourham, Committee Member
The 48Ti(p, gamma)49V reaction was studied in the energy range Ep = 2.0--2.8 MeV. This research was performed at the High Resolution Laboratory at Triangle Universities Nuclear Laboratory. Capture spectra were obtained for 27 compound nuclear resonances whose resonance parameters (energy, spin, parity, and particle widths) had been determined in earlier experiments. The gamma rays were measured with a Compton-suppressed HPGe detector. A total of ten 1/2+ resonances, nine 1/2- resonances and eight 3/2 - resonances were studied. The primary motivation for these measurements was to provide a data set suitable for testing the applicability of conventional statistical models in this mass region. A secondary motivation was to evaluate the 'low level population' method (which works very well in heavier nuclei) in this mass region. The relative populations of 10 of the first 12 excited states of 49V were determined (the relative populations of two of the low-lying states could not be obtained due to gamma-ray resolution limitations). The average relative populations were similar for the three types of resonances. The experimental relative populations were compared with calculations performed with the statistical model code DICEBOX. Results for a variety of radiative strength function models yielded qualitatively similar results. These simulations agree only qualitatively with the experimental relative populations. Although the simulations suggest that the low level population method may work in this nuclide, the experimental results are at best ambiguous. Thus the statistical approach provides a reasonable qualitative description (but not a quantitative description) of the relative population of low-lying states in 49V.
Reactor Loose Part Damage Assessments on Steam Generator Tube Sheets
(2010-03-15) Proctor, William Cyrus; Moody T. Chu, Committee Member; Mohamed A. Bourham, Committee Member; J. Michael Doster, Committee Chair
PROCTOR, WILLIAM CYRUS. Reactor Loose Part Damage Assessments on Steam Generator Tube Sheets. (Under the direction of Joseph Michael Doster). Damage from loose parts inside reactor systems can potentially cause integrity issues that jeopardize the operations of these facilities. Parts such as nuts, bolts, pins, sections of tubing and even hand tools are found inside the primary circuits of PWRs [Michel]. These parts carried by the coolant flow impact structures including the steam generator tube sheets and can cause significant damage leading to the unscheduled shut down of a facility. In this work we assess the behaviors of typical loose parts that may reside in the primary coolant system. Validations of scaled simulations are linked to previous experiments conducted by Shi [Shi]. Monte Carlo simulations of typical impact and energy distributions on a representative steam generator are analyzed and discussed. To obtain a more complete understanding of loose part damage caused to the tube sheet of PWR steam generators, CFD using the ANSYS CFX software package is used to compute detailed three dimensional flow fields within the steam generator inlet plenum. The flow field information is then input into a Monte Carlo program developed as part of this work to predict the trajectory of the loose part. Existing software packages lack the ability to track finite volume, finite mass particles. Additionally, there were no packages available that allowed for detailed manipulation of the collision physics necessary to accurately model impacts. The particle tracking program developed here then allows for the calculation of loose part impact locations and the energy imparted from loose part impacts with the tube sheet surface. Ultimately given this information along with the previous models developed by Shi, damage rates can be estimated aiding in the development of guidelines to improve the decision making process when loose parts are detected in the primary coolant system. As part of previous research, a 1:8 scaled model of the McGuire steam generator inlet plenum and tube sheet was constructed by Shi. This scaled steam generator tube sheet impact pattern experiment was run with two different types of hexagonal nuts and varied fluid inlet velocities. These experiments serve as a benchmark reference for development of the computational models in this work. Simulations of a full scale system similar to that of a Westinghouse Model D steam generator have also been performed. Detailed impact analysis is conducted as a function of coolant temperature, coolant inlet velocity, loose part type, shape, mass, density, initial starting location and initial kinetic energy. No a priori knowledge is assumed for the initial starting location and initial kinetic energy of the parts. Full scale results are compared to the scaled experiment to assess the validity of making predictions using only a scaled simulation.
A Study on the Feasibility of Electron-based Accelerator Driven Systems for Nuclear Waste Transmutation
(2006-08-07) Liu, Yaxi; David McNelis, Committee Co-Chair; Man-sung Yim, Committee Chair; Mohamed A. Bourham, Committee Member; John F. Muth, Committee Member
Nuclear waste transmutation is an important option for the development of advanced fuel cycle and effective nuclear waste management. The electron accelerator driven system (ADS) was investigated in the study for nuclear waste transmutation as an alternative to proton based ADS. Target design and optimization was carried out to obtain maximum neutron generation. Subcritical core design based on single and multiple targets was investigated. System performance between electron-based ADS and proton-based ADS was compared in terms of neutron generation rate, transmutation efficiency and power generation. It was determined that the electron-based target was capable of providing high neutron flux, small target geometry size, small scale subcritical core, and low radiation damage. Multiple target design in the electron-driven ADS was also explored to flatten power distribution in the ADS subcritical core. Regarding transmutation, the power peaking factors in both the electron- and proton- ADS increase ~ 10% during the burnup period of 700 days. Thermal power in proton ADS is higher than that of electron ADS by a factor ~ 20. The transmutation effectiveness of preliminary electron-based ADS is smaller by a factor of 11 compared to preliminary proton-based ADS. Proton ADS has higher radiation damage to target materials and surrounding materials. The capital cost for electron-based and proton-based accelerator facility is fairly comparable with the cost of proton-based facilities being slightly higher by a factor of 20%. Comparing with the proton-driven ADS, the electron-driven ADS pros include small target size and small core scale, multiple target possibility for low PPF, low radiation damage to target surroundings, wide availability electron beam at ~100 MeV, and low capital cost of electron accelerator facility. There are also aspects against electron-driven ADS, including low efficiency of neutron generation rate, low transmutation efficiency, low thermal power, and electricity generation.
Thermal Design of Wide Beam Area X-Ray Sources
(2009-03-13) Bobolea, Nicolae Alin; J. Michael Doster, Committee Chair; K. Linga Murty, Committee Member; Mohamed A. Bourham, Committee Member
Diffraction Enhanced Imaging (DEI) with x-ray radiation provided by a synchrotron source has been shown to provide good image contrast at lower radiation dose for materials with small x-ray attenuation coefficient As a result, DEI has received significant interest for digital mammography and other medical imaging applications. However, deployment of a synchrotron source at a medical facility is not currently feasible due to its size and costs. Consequently, a compact x-ray source capable of delivering x-ray intensities and beam collimation similar to a synchrotron accelerator is desirable. A wide beam area x-ray source has been suggested as a possible alternative to a synchrotron source, with the x-rays generated by electron bombardment of a suitable target material. Previous research work demonstrated a prototype scale cylindrical shaped oxygen free copper target with a layer of molybdenum to be feasible from an engineering perspective. An industrial size DEI facility requires a scale-up of the proof-of-principle design. The x-ray flux necessary for high image quality implies significant heat loading on the x-ray source. Safe operation of a full scale DEI facility is reliant upon a thermal management solution capable of rejecting this heat. An active target cooling system has been proposed and its performance has been evaluated through CFD simulation. The design ensures the maximum target temperature is maintained at reasonable levels and coolant boiling is not reached under the most demanding operating conditions.