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FullText Articles in Engineering
TimeOfFlight And Energy Loss Analysis On The Unm Fission Fragment Spectrometer, Shelby Fellows
TimeOfFlight And Energy Loss Analysis On The Unm Fission Fragment Spectrometer, Shelby Fellows
Nuclear Engineering ETDs
The University of New Mexico spectrometer experimental work has been used to provide an eventbyevent fission product measurement to aid in filling in the gaps in existing fission product yield data, as part of the Los Alamos National Lab Spectrometer for Ion Detection in Fission Research project (SPIDER) collaboration. This thesis examines the timeofflight (TOF) component of the spectrometer towards improving the resolution of the system. Different thicknesses of TOF conversion foils were examined with alpha particles and fission fragments: 20, 55, and 100 µg/cm^{2 }carbon foils. For the thinnest carbon foil studied, a timing resolution of 160 ...
Impact Of Fuel Rod Coatings On Reactor Performance And Safety, Ian Robert Stewart
Impact Of Fuel Rod Coatings On Reactor Performance And Safety, Ian Robert Stewart
Masters Theses
This study evaluates the use of a ceramic coating on the Zralloy cladding within a PWR using four ceramic compounds of 5 and 10 micron thicknesses: ZrO2, TiAlN, Ti2AlC, and Ti3AlC2. The film’s impact is assessed for variation on: reactivity, fuel cycle length, maximum temperature, film’s roughness, and transient conditions. The reactivity is analyzed using the following methods: change in the multiplication factor (k) each film introduces to the system using the ABH method, and Monte Carlo software (MCNP). Both methods are in good agreement, yielding less than half a percent change from a reference, nofilm fuel pin ...
Improvement Of Spent Fuel Storage With Advanced Mechanical Shielding Placement, Gordon M. Petersen, Nicole Galante, Hannah Hale, Dylan Richardson, Robert Vance
Improvement Of Spent Fuel Storage With Advanced Mechanical Shielding Placement, Gordon M. Petersen, Nicole Galante, Hannah Hale, Dylan Richardson, Robert Vance
Chancellor’s Honors Program Projects
No abstract provided.
Modified Sodium Diuranate Process For The Recovery Of Uranium From Uranium Hexafluoride Transport Cylinder Wash Solution, Austin Dean Meredith
Modified Sodium Diuranate Process For The Recovery Of Uranium From Uranium Hexafluoride Transport Cylinder Wash Solution, Austin Dean Meredith
Theses and Dissertations
Uranium hexafluoride (UF_{6}) containment cylinders must be emptied and washed every five years in order to undergo recertification, according to ANSI standards. During the emptying of the UF_{6} from the cylinders, a thin residue, or heel, of UF_{6} is left behind. This heel must be removed in order for recertification to take place.
To remove it, the inside of the containment cylinder is washed with acid and the resulting solution generally contains three or four kilograms of uranium. Thus, before the liquid solution can be disposed of, the uranium must be separated. A modified sodium diuranate (SDU ...
Advanced Fuels Modeling: Evaluating The SteadyState Performance Of Carbide Fuel In HeliumCooled Reactors Using Frapcon 3.4, Luke H. Hallman
Advanced Fuels Modeling: Evaluating The SteadyState Performance Of Carbide Fuel In HeliumCooled Reactors Using Frapcon 3.4, Luke H. Hallman
Theses and Dissertations
Uranium carbide (UC) has long been considered a potential alternative to uranium dioxide (UO2) fuel, especially in the context of Gen IV gascooled reactors. It has shown promise because of its high uranium density, good irradiation stability, and especially high thermal conductivity. Despite its many benefits, UC is known to swell at a rate twice that of UO2. However, the swelling phenomenon is not well understood, and we are limited to a weak empirical understanding of the swelling mechanism.
One suggested cladding for UC is silicon carbide (SiC), a ceramic that demonstrates a number of desirable properties. Among them are ...
Total System Performance Assessment Of The Proposed High Level Radioactive Waste Repository Site At Genting Island, Karimunjawa, Indonesia , Yudi Utomo Imardjoko
Total System Performance Assessment Of The Proposed High Level Radioactive Waste Repository Site At Genting Island, Karimunjawa, Indonesia , Yudi Utomo Imardjoko
Retrospective Theses and Dissertations
Indonesia is about to enter the nuclear era by building several nuclear power plants in the near future. Numerous issues must be addressed to evaluate the impact on the environment of highlevel radioactive waste (HLRW) generated by these plants. This consideration of HLRW management must be done in a timely manner;This dissertation discusses the methodology for the Genting repository site development plan. The approach to the development of the repository is divided into three main areas, the inventory of HLRW, the barrier systems (natural and engineered), and the physical conditions of the site. The radionuclide inventory and waste form ...
Modeling XRay Scattering Process And Applications Of The Scattering Model , Taher Lutfi AlJundi
Modeling XRay Scattering Process And Applications Of The Scattering Model , Taher Lutfi AlJundi
Retrospective Theses and Dissertations
Computer modeling of nondestructive inspections with xrays is proving to be a very useful tool for enhancing the performance of these techniques. Two xray based inspection techniques are considered in this study. The first is "Radiographic Inspection", where an existing simulation model has been improved to account for scattered radiation effects. The second technique is "Inspection with Compton backscattering", where a new simulation model has been developed;The effect of scattered radiation on a simulated radiographic image can be insignificant, equally important, or more important than the effect of the uncollided flux. Techniques to account for the scattered radiation effects ...
Performance Assessment Modeling Of Alternative High Level Nuclear Wasteforms , William Mark Nutt
Performance Assessment Modeling Of Alternative High Level Nuclear Wasteforms , William Mark Nutt
Retrospective Theses and Dissertations
Performance assessment (PA) analyses have been completed for high level nuclear waste forms derived from once through Light Water Reactor (LWR) spent fuel (SF), defense high level waste (DHLW), the closed advanced liquid metal reactor (ALMR) pyroprocess fuel cycle, and the front end processing of LWR spent fuel for possible recycling of the actinides in the closed ALMR fuel cycle. The IMARC (Integrated Multiple Assumptions and Release Calculations) and RIP (Repository Integration Program) PA tools were utilized to predict the performance of these wasteforms for emplacement in the proposed Yucca Mountain Repository;A model was developed to predict the time ...
Nuclear Plant Diagnostics Using Neural Networks With Dynamic Input Selection , Anujit Basu
Nuclear Plant Diagnostics Using Neural Networks With Dynamic Input Selection , Anujit Basu
Retrospective Theses and Dissertations
The work presented in this dissertation explores the design and development of a large scale nuclear power plant (NPP) fault diagnostic system based on artificial neural networks (ANNs). The viability of detecting a large number of transients in a NPP using ANNs is demonstrated. A new adviser design is subsequently presented where the diagnostic task is divided into component parts, and each part is solved by an individual ANN. This new design allows the expansion of the diagnostic capabilities of an existing adviser by modifying the existing ANNs and adding new ANNs to the adviser;This dissertation also presents an ...
Reliability Assessment Of Nuclear Power Plant FaultDiagnostic Systems Using Artificial Neural Networks , Keehoon Kim
Reliability Assessment Of Nuclear Power Plant FaultDiagnostic Systems Using Artificial Neural Networks , Keehoon Kim
Retrospective Theses and Dissertations
The assurance of the diagnosis obtained from a nuclear power plant (NPP) faultdiagnostic advisor based on artificial neural networks (ANNs) is essential for the practical implementation of the advisor to transient detection and identification. The objectives of this study are to develop a validation and verification technique suitable for ANNs and apply it to the faultdiagnostic advisor. The validation and verification is realized by estimating error bounds on the advisor's diagnoses. The two different partition criteria are developed to create computationally effective partitions for generating the error information associated with the advisor performance. The bootstap partition criterion (BPC) and ...
Knowledge Base Expert System Control Of Spatial Xenon Oscillations In Pressurized Water Reactors , Serhat Alten
Knowledge Base Expert System Control Of Spatial Xenon Oscillations In Pressurized Water Reactors , Serhat Alten
Retrospective Theses and Dissertations
Nuclear reactor operators are required to pay special attention to spatial xenon oscillations during the loadfollow operation of pressurized water reactors. They are expected to observe the axial offset of the core, and to estimate the correct time and amount of necessary control action based on heuristic rules given in axial offset control strategies;Current control methods of axial xenon oscillations are knowledge intensive, and heuristic in nature. An expert system, ACES (Axial offset Control using Expert Systems) is developed to implement a heuristic constant axial offset control procedure to aid reactor operators in increasing the plant reliability by reducing ...
Neutron And Proton Dosimetry At The Lampf 800Mev Proton Accelerator , Dorothy R. Davidson
Neutron And Proton Dosimetry At The Lampf 800Mev Proton Accelerator , Dorothy R. Davidson
Retrospective Theses and Dissertations
Characterization of the radiation environment at the Los Alamos Spallation Radiation Effects Facility (LASREF) has been completed. Monte Carlo neutronics calculations, foil activation experiments, and radiation damage calculations were performed to determine the neutron and proton flux and energy spectrum at the LASREF;Thirteen foil activation experiments using highpurity iron, scandium, copper, nickel, vanadium, titanium, aluminum and cobalt foils were irradiated in a "rabbit" dosimetry insert to measure the neutron and proton flux and energy spectrum in neutron irradiation ports outside the IP targets and the beam stop and one proton irradiation port. Cobalt, scandium, and iron foils wrapped with ...
Incorporation Of A Rectangular Void Into A ThreeDimensional Neutron Diffusion Nodal Model , Mohamed Boussoufi
Incorporation Of A Rectangular Void Into A ThreeDimensional Neutron Diffusion Nodal Model , Mohamed Boussoufi
Retrospective Theses and Dissertations
The problem of a parallelipiped void occupying a nodal position in the diffusion nodal method is investigated. To solve for the neutron partial currents that couple the void to its surrounding nodes, the concept of overall transfer matrix is introduced. This transfer matrix, when multiplied by the column vector containing the Legendre coefficients of the spatial flux in the neighboring node, produces the outbound neutron partial currents. This method does not require the solution of the neutron transport or diffusion equation inside the void but rather uses the faces of the void as P1 distributed neutron sources and finds the ...
Design Of An Expert System For Failed Fuel Identification And Surveillance In EbrIi , Ramin Mikaili
Design Of An Expert System For Failed Fuel Identification And Surveillance In EbrIi , Ramin Mikaili
Retrospective Theses and Dissertations
Since 1977, a program has been underway at experimental breeder reactorII (EBRII) to evaluate the performance of metal and mixedoxide liquid metal reactor (LMR) fuel elements after clad failure (breach). The motivation for this activity, called run beyond cladding breach (RBCB) testing, is to continue safe operation of EBRII after occurrence of a single or multiple clad failures. Principal safety concerns are excessive release of fission gas (FG) and mixedoxide fuel/sodium interaction, which result in release of byproducts to the sodium and may result in blockage of coolant flow. Excessive release of FG is controlled by the cover gas ...
Thermal Hydraulic Aspects Of An Unconventional Liquid Metal Reactor , Cemal Niyazi Sökmen
Thermal Hydraulic Aspects Of An Unconventional Liquid Metal Reactor , Cemal Niyazi Sökmen
Retrospective Theses and Dissertations
The Trench Reactor (TR) is a liquid sodium cooled fast power reactor. The reactor core is fueled with UPuZr metal fuel and generates 800 MW of thermal power. The core is located in a sodium pool which is contained in a thin and deep rectangular vessel. Also located in the pool are the two intermediate heat exchangers (IHX) and the two primary pumps. The liquid sodium exists the core at 485°C and is pumped through the IHXs where it is cooled to 343°C and enters the core through an inlet plenum. The reactor building atmosphere is nitrogen which ...
Computer Simulation Of Plastic Deformation In Irradiated Metals , Üner Çolak
Computer Simulation Of Plastic Deformation In Irradiated Metals , Üner Çolak
Retrospective Theses and Dissertations
A computerbased model is developed for the localized plastic deformation in irradiated metals by dislocation channeling, and it is applied to irradiated single crystals of niobium. In the model, the concentrated plastic deformation in the dislocation channels is postulated to occur by virtue of the motion of dislocations in a series of piletips on closely spaced parallel slip planes. The dynamics of this dislocation motion is governed by an experimentally determined dependence of dislocation velocity on shear stress. This leads to a set of coupled differential equations for the positions of the individual dislocations in the pileup as a function ...
Stochastic Memory Process And Its Application To Cumulative Outage Time In Nuclear Power Plants , Mohamad Ali Azarm
Stochastic Memory Process And Its Application To Cumulative Outage Time In Nuclear Power Plants , Mohamad Ali Azarm
Retrospective Theses and Dissertations
The safety performance of operating nuclear power plants is strongly affected by the unavailability of safety systems that are designed to mitigate accident conditions. The unavailability of these safety systems during plant operation is controlled by the plant's technical specifications which prescribes limits on the downtime duration (outage time) of the individual safety equipment. In this study, risk and reliabilitybased methodologies for the determination of allowable cumulative downtime for safety components and safety systems are developed. The limits on the cumulative downtime durations are determined by taking into account the statistical variations expected from a stochastic process which models ...
A Modular Nodal Method For Solving The Neutron Transport Equation Using Spherical Harmonics In Two Dimensions , Feyzi Inanc
A Modular Nodal Method For Solving The Neutron Transport Equation Using Spherical Harmonics In Two Dimensions , Feyzi Inanc
Retrospective Theses and Dissertations
A modular nodal method is developed for solving the neutron transport equation by using the spherical harmonics approximation in two dimensional Cartesian coordinates. The spherical harmonics approximation is based upon the second order evenparity form of the neutron transport equation. The boundary conditions of the spherical harmonics approximation are manipulated to have the forms analogous to the partial currents in the neutron diffusion equation. The relationships are developed for generating both the second order spherical harmonic equations and the boundary conditions in an automatic manner. The nodal method developed is based upon a least squares minimization technique. In that method ...
The Sk(N) Approximation: A New Method For Solving Integral Transport Equations , Zekeriya Altaç
The Sk(N) Approximation: A New Method For Solving Integral Transport Equations , Zekeriya Altaç
Retrospective Theses and Dissertations
A high order transport approximation, the SK[subscript]N approximation, a mnemonic for "synthetic kernel", is suggested for solving the integral transport equation. The method relies on approximating the integral transport kernels by a sum of diffusionlike kernels. The integral equation is then reducible to a set of coupled differential equations, and the boundary conditions (black body and reflecting boundary conditions) are established. These equations, the SK[subscript]N equations, are solved for benchmark problems. The benchmark problems include one and twodimensional homogeneous and heterogeneous cell configurations found in nuclear reactor applications. The solutions of the benchmark problems are compared ...
Reprocessing Of LongCooled Nuclear Fuel: Process Description And Plant Design , Okan H. Zabunoglu
Reprocessing Of LongCooled Nuclear Fuel: Process Description And Plant Design , Okan H. Zabunoglu
Retrospective Theses and Dissertations
Purex reprocessing of 10year cooled LWR fuel was investigated. The three major areas of research were (1) process description and flowsheet calculations, (2) plant design, and (3) cost estimation. Involvement of nearly 13 times less radioactivity in the 10year cooled fuel than that in the 150day (standard cooling time) cooled fuel, and the adoption of coprocessing of U and Pu in a new flowsheet rather than the complete partitioning of the standard purex method resulted in several simplifications in process and design over the standard purex method handling 150day cooled fuel, leading to a simpler and more economic design for ...
Implementation Of An Expert System For Xenon Spatial Control In Pressurized Water Reactors , SunKyo Chung
Implementation Of An Expert System For Xenon Spatial Control In Pressurized Water Reactors , SunKyo Chung
Retrospective Theses and Dissertations
Most commercial pressurized water reactors are unstable to xenon oscillations that occur during loadfollow operation;Control of the axial xenon oscillations is a knowledge and experienceintensive activity for reactor operators. To aid reactor operators in the control of axial xenon oscillations, an advisory expert system was developed;A rulebased expert system shell, INSIGHT2+, was used to build the expert system which was interfaced with a microcomputerbased core control model of a pressurized water reactor, graphic engine, and data base;A core control model described by onegroup diffusion theory with moderator temperature and xenon feedbacks was used to develop heuristic control ...
Development Of TwoGroup, TwoDimensional, Frequency Dependent Detector Adjoint Function Based On The Nodal Method , Soli T. Khericha
Development Of TwoGroup, TwoDimensional, Frequency Dependent Detector Adjoint Function Based On The Nodal Method , Soli T. Khericha
Retrospective Theses and Dissertations
A concept of local/global components, based on the frequency dependent detector adjoint function, and a nodalization technique was utilized in the development of one and two dimensional computer codes to calculate the response of a detector to a vibrating absorber in reactor cores. The frequency dependent detector adjoint functions presented by complex equations were expanded into real and imaginary parts. In the nodalization technique, the flux is expanded into polynomials about the center point of each node;The phase angles and the magnitudes of the twoenergy group detector adjoint functions were calculated for a neutron detector located in the ...
A ThreeDimensional Nodal Solution For The Frequency Dependent Neutron Diffusion Equation , Abdulghani M. Melaibari
A ThreeDimensional Nodal Solution For The Frequency Dependent Neutron Diffusion Equation , Abdulghani M. Melaibari
Retrospective Theses and Dissertations
This research involves the development of a threedimensional nodal code that calculates the Fourier transformed regular or adjoint neutron flux for a nuclear reactor. This numerical technique can be used in the nuclear reactor noise analysis field for identifying and locating vibrating reactor core components;The mathematical equations were developed and two types of solutions were obtained. The first solution was a modification of a threedimensional nodal model developed to handle multigroup neutron diffusion equations. In this model, the Fourier transformed fluxes were expanded in the Legendre polynomial form. The second is an analytical procedure developed for a simple geometry ...
Disposal Of Spent Nuclear Fuel And HighLevel Waste: Design And Technical/Economic Analysis , Jordi RoglansRibas
Disposal Of Spent Nuclear Fuel And HighLevel Waste: Design And Technical/Economic Analysis , Jordi RoglansRibas
Retrospective Theses and Dissertations
An economic model for the back end of the nuclear fuel cycle was developed for a oncethrough cycle, a standard reprocessing cycle, and a reprocessing cycle with fractionation of cesium and strontium. The development of the model was performed under the expected political constraints and scenario for the first nuclear waste repository. Technical issues concerning the repository design were analyzed, in particular the thermal design. A parametric thermal analysis was performed for waste emplaced in five different geologic formations: salt, granite, basalt, shallow tuff, and deep tuff. The results of the thermal analysis, in the form of maximum permissible loadings ...
Vibration Identification Of Nuclear Reactor Components By Statistical Analysis Of Neutron Noise , John Thomas Sankoorikal
Vibration Identification Of Nuclear Reactor Components By Statistical Analysis Of Neutron Noise , John Thomas Sankoorikal
Retrospective Theses and Dissertations
The possibility of identifying vibrating components, in a nuclear reactor core, through the use of statistical techniques has been investigated. Mechanical vibrations produce neutron noise which appears as fluctuations in detector signals. Theory pertaining to the production of neutron noise is discussed. Vibrations are characterized by location and vibration trajectory parameters. Maximumlikelihood and confidenceregion techniques were developed to estimate these parameters. Computer experiments were carried out using simulated detector signals for a simplified reactor model. The sensitivity of the techniques was investigated by parametrically studying the effects of noise level in the detector signal and the presence of external noise ...
Formulation And Analysis Of Higher Order Finite Difference Approximations To The Neutron Diffusion Equation , Mohammed Benghanem
Formulation And Analysis Of Higher Order Finite Difference Approximations To The Neutron Diffusion Equation , Mohammed Benghanem
Retrospective Theses and Dissertations
Analyses have been made of the truncation error for the following finite difference approximations to the eigenvalue and boundary value problems evolving from the onegroup neutron diffusion equation: (i) The sevenpoint relation; (ii) The fifteenpoint relation; (iii) The nineteenpoint relation; and (iv) The twentyseven point relation. These methods have been derived using a Taylor series expansion technique and applied to the Laplacian operator contained in that equation in (x,y,z) geometry for various reactor configurations and boundary conditions;It has been shown that for methods ii and iii, a 4th order truncation error can be achieved, whereas for the ...
Sensitivity Analysis Approach For Robust Probabilistic Risk Assessment , Shahid Ahmed
Sensitivity Analysis Approach For Robust Probabilistic Risk Assessment , Shahid Ahmed
Retrospective Theses and Dissertations
The main objective of this investigation is to develop a robust and simplified Probabilistic Risk Assessment (PRA) approach specifically oriented to produce results for risk management decisions of high technology systems. The techniques are based on defining a set of three Significant Indices which quantify the importance of each component, and hence develop a ranking of the components, both with respect to the mean and variance, in the fault tree/event tree structure. The variations in the Top Event probability distribution upon the variations in the component input probability distributions are also evaluated, as well as the first and second ...
Development Of A Polynomial Nodal Model To The Multigroup Transport Equation In One Dimension , Masoud Feiz
Development Of A Polynomial Nodal Model To The Multigroup Transport Equation In One Dimension , Masoud Feiz
Retrospective Theses and Dissertations
A polynomial nodal model which uses Legendre polynomial expansions was developed for the multigroup transport equation in one dimension. The development depends upon the least squares minimization of the residuals using the approximate functions over the node. Analytical expressions were developed for the polynomial coefficients. The odd moments of the angular neutron flux over the half ranges were used at the internal interfaces, and the Marshak boundary condition was used at the external boundaries. Sample problems with fine mesh finite difference solutions of the diffusion and transport equations were used for comparison with the model.
Energy Balance Calculations For A Multipole Target Plasma Fusion Reactor , Terry Edwin Dix
Energy Balance Calculations For A Multipole Target Plasma Fusion Reactor , Terry Edwin Dix
Retrospective Theses and Dissertations
Thermonuclear fusion reactors have not yet achieved breakeven; high plasma temperatures are required to obtain high reaction rates. Accompanying the high plasma temperatures are high bremsstrahlung radiation losses, plasma instabilities, first wall problems and large amounts of energy for plasma containment. To reduce the detrimental effects and maintain high reaction rates, a two component target plasma system was proposed with one of the fuel species acting as a target plasma magnetically confined at a relatively low temperature. The second fuel species is then injected at high energy into the target plasma to interact with the confined plasma as it slows ...
Reliability Analysis For The Emergency Power System Of A Pressurized Water Reactor Facility During A Loss Of Offsite Power Transient , SeeMeng Wong
Reliability Analysis For The Emergency Power System Of A Pressurized Water Reactor Facility During A Loss Of Offsite Power Transient , SeeMeng Wong
Retrospective Theses and Dissertations
The anticipated transient involving loss of offsite power has been identified as a major potential contributor of risk for a pressurized water reactor power generating facility. The capability of the emergency power system to supply adequate power to engineering safety systems is important for limiting the progression of the transient event and avoiding reactor core degradation;The evaluation of emergency power system reliability, using probabilistic risk assessment techniques with updated failure data, shows that the unavailability of this system is significantly higher than had been assessed earlier. The significant contributors to system unavailability are diesel generator operability problems, particularly when ...