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Full-Text Articles in Nuclear Engineering

Total System Performance Assessment Of The Proposed High Level Radioactive Waste Repository Site At Genting Island, Karimunjawa, Indonesia , Yudi Utomo Imardjoko Jan 1995

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 high-level 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 ...


Nuclear Plant Diagnostics Using Neural Networks With Dynamic Input Selection , Anujit Basu Jan 1995

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 ...


Modeling X-Ray Scattering Process And Applications Of The Scattering Model , Taher Lutfi Al-Jundi Jan 1995

Modeling X-Ray Scattering Process And Applications Of The Scattering Model , Taher Lutfi Al-Jundi

Retrospective Theses and Dissertations

Computer modeling of nondestructive inspections with x-rays is proving to be a very useful tool for enhancing the performance of these techniques. Two x-ray 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 Jan 1995

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 ...


Reliability Assessment Of Nuclear Power Plant Fault-Diagnostic Systems Using Artificial Neural Networks , Keehoon Kim Jan 1994

Reliability Assessment Of Nuclear Power Plant Fault-Diagnostic Systems Using Artificial Neural Networks , Keehoon Kim

Retrospective Theses and Dissertations

The assurance of the diagnosis obtained from a nuclear power plant (NPP) fault-diagnostic 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 fault-diagnostic 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 Jan 1992

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 load-follow 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 ...


Design Of An Expert System For Failed Fuel Identification And Surveillance In Ebr-Ii , Ramin Mikaili Jan 1990

Design Of An Expert System For Failed Fuel Identification And Surveillance In Ebr-Ii , Ramin Mikaili

Retrospective Theses and Dissertations

Since 1977, a program has been underway at experimental breeder reactor-II (EBR-II) to evaluate the performance of metal and mixed-oxide 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 EBR-II after occurrence of a single or multiple clad failures. Principal safety concerns are excessive release of fission gas (FG) and mixed-oxide fuel/sodium interaction, which result in release of by-products to the sodium and may result in blockage of coolant flow. Excessive release of FG is controlled by the cover gas ...


Neutron And Proton Dosimetry At The Lampf 800-Mev Proton Accelerator , Dorothy R. Davidson Jan 1990

Neutron And Proton Dosimetry At The Lampf 800-Mev 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 high-purity 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 Three-Dimensional Neutron Diffusion Nodal Model , Mohamed Boussoufi Jan 1990

Incorporation Of A Rectangular Void Into A Three-Dimensional 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 ...


Stochastic Memory Process And Its Application To Cumulative Outage Time In Nuclear Power Plants , Mohamad Ali Azarm Jan 1989

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 reliability-based 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 Jan 1989

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 even-parity 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 ...


Thermal Hydraulic Aspects Of An Unconventional Liquid Metal Reactor , Cemal Niyazi Sökmen Jan 1989

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 U-Pu-Zr 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 ...


The Sk(N) Approximation: A New Method For Solving Integral Transport Equations , Zekeriya Altaç Jan 1989

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 diffusion-like 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 two-dimensional homogeneous and heterogeneous cell configurations found in nuclear reactor applications. The solutions of the benchmark problems are compared ...


Computer Simulation Of Plastic Deformation In Irradiated Metals , Üner Çolak Jan 1989

Computer Simulation Of Plastic Deformation In Irradiated Metals , Üner Çolak

Retrospective Theses and Dissertations

A computer-based 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 pile-tips 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 pile-up as a function ...


Reprocessing Of Long-Cooled Nuclear Fuel: Process Description And Plant Design , Okan H. Zabunoglu Jan 1988

Reprocessing Of Long-Cooled Nuclear Fuel: Process Description And Plant Design , Okan H. Zabunoglu

Retrospective Theses and Dissertations

Purex reprocessing of 10-year 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 10-year cooled fuel than that in the 150-day (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 150-day cooled fuel, leading to a simpler and more economic design for ...


Implementation Of An Expert System For Xenon Spatial Control In Pressurized Water Reactors , Sun-Kyo Chung Jan 1988

Implementation Of An Expert System For Xenon Spatial Control In Pressurized Water Reactors , Sun-Kyo Chung

Retrospective Theses and Dissertations

Most commercial pressurized water reactors are unstable to xenon oscillations that occur during load-follow operation;Control of the axial xenon oscillations is a knowledge- and experience-intensive activity for reactor operators. To aid reactor operators in the control of axial xenon oscillations, an advisory expert system was developed;A rule-based expert system shell, INSIGHT2+, was used to build the expert system which was interfaced with a microcomputer-based core control model of a pressurized water reactor, graphic engine, and data base;A core control model described by one-group diffusion theory with moderator temperature and xenon feedbacks was used to develop heuristic control ...


Development Of Two-Group, Two-Dimensional, Frequency Dependent Detector Adjoint Function Based On The Nodal Method , Soli T. Khericha Jan 1987

Development Of Two-Group, Two-Dimensional, 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 two-energy group detector adjoint functions were calculated for a neutron detector located in the ...


A Three-Dimensional Nodal Solution For The Frequency Dependent Neutron Diffusion Equation , Abdulghani M. Melaibari Jan 1987

A Three-Dimensional Nodal Solution For The Frequency Dependent Neutron Diffusion Equation , Abdulghani M. Melaibari

Retrospective Theses and Dissertations

This research involves the development of a three-dimensional 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 three-dimensional 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 High-Level Waste: Design And Technical/Economic Analysis , Jordi Roglans-Ribas Jan 1987

Disposal Of Spent Nuclear Fuel And High-Level Waste: Design And Technical/Economic Analysis , Jordi Roglans-Ribas

Retrospective Theses and Dissertations

An economic model for the back end of the nuclear fuel cycle was developed for a once-through 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 ...


Sensitivity Analysis Approach For Robust Probabilistic Risk Assessment , Shahid Ahmed Jan 1986

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 Jan 1986

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.


Vibration Identification Of Nuclear Reactor Components By Statistical Analysis Of Neutron Noise , John Thomas Sankoorikal Jan 1986

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. Maximum-likelihood and confidence-region 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 Jan 1986

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 one-group neutron diffusion equation: (i) The seven-point relation; (ii) The fifteen-point relation; (iii) The nineteen-point relation; and (iv) The twenty-seven 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 ...


Energy Balance Calculations For A Multipole Target Plasma Fusion Reactor , Terry Edwin Dix Jan 1985

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 , See-Meng Wong Jan 1984

Reliability Analysis For The Emergency Power System Of A Pressurized Water Reactor Facility During A Loss Of Offsite Power Transient , See-Meng 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 ...


Quantitative Assessment Of Human Contribution To Risk In Nuclear Power Plants , Abdallah Ahmad Ezzedin Jan 1983

Quantitative Assessment Of Human Contribution To Risk In Nuclear Power Plants , Abdallah Ahmad Ezzedin

Retrospective Theses and Dissertations

It is known from operation history of nuclear power plants (NPPs) that human factors considerations and human reliability analysis (HRA) are necessary for complete safety analysis of NPPs. HRA is usually performed to estimate the influence of human errors on the unavailability of various safety systems of NPPs;The objective of this work is to assess quantitatively the human contribution to the unavailability of those safety systems involved in the S(,2)C accident sequence. S(,2)C sequence is defined as a small loss of coolant accident (LOCA) and core meltdown after containment failure. This sequence is selected for ...


The Effect Of Rare-Earth Element Additions On Microstructural Properties And Irradiation Behavior Of An Fe-Ni-Cr Alloy For Lmfbr And Fusion Reactor Applications , Jin-Young Park Jan 1983

The Effect Of Rare-Earth Element Additions On Microstructural Properties And Irradiation Behavior Of An Fe-Ni-Cr Alloy For Lmfbr And Fusion Reactor Applications , Jin-Young Park

Retrospective Theses and Dissertations

This study consists of a survey of the effect of yttrium, lanthanum, and cerium rare-earth additions on the microstructure and radiation swelling behavior of an Fe-25.6Ni-8.7Cr-3.3Ti-1.6Al alloy. The undoped alloy was investigated in the as-received, annealed, and arc-melted conditions, and twelve arc melted and rare-earth doped alloys were prepared (doping levels of 0.05, 0.1, 0.5, and 1.0 wt % for each of the three rare earths). The ion bombardments were carried out at 570 and 600(DEGREES)C with 4 MeV Ni or Fe ions to nominal 100 dpa and to 100 and ...


Analysis Of The Nine-Point Finite Difference Approximation For The Heat Conduction Equation In A Nuclear Fuel Element , Mohamed Kadri Jan 1983

Analysis Of The Nine-Point Finite Difference Approximation For The Heat Conduction Equation In A Nuclear Fuel Element , Mohamed Kadri

Retrospective Theses and Dissertations

The time dependent heat conduction equation in the x-y Cartesian geometry is formulated in terms of a nine-point finite difference relation using a Taylor series expansion technique. The accuracy of the nine-point formulation over the five-point formulation has been tested and evaluated for various reactor fuel-cladding plate configurations using a computer program. The results have been checked against analytical solutions for various model problems;The following cases were considered in the steady-state condition: (a) The thermal conductivity and the heat generation were uniform. (b) The thermal conductivity was constant, the heat generation variable. (c) The thermal conductivity varied linearly with ...


Development And Application Of A Decision Methodology For The Planning Of Nuclear Research And Development In Saudi Arabia , Waleed Hussain Abulfaraj Jan 1983

Development And Application Of A Decision Methodology For The Planning Of Nuclear Research And Development In Saudi Arabia , Waleed Hussain Abulfaraj

Retrospective Theses and Dissertations

The present study involves adapting two formal decision methodologies to the selection of alternative nuclear energy strategies. Multiattribute utility theory and fuzzy set theory are selected to accommodate for decision makers' preferences and for imprecisions in evaluation of factors impacting a decision, respectively;Multiattribute Utility theory (MAU) is here employed to evaluate four appropriate research reactor facilities to determine the optimal choice in order to meet the needs of Saudi Arabia. These facilities are similar to University of Michigan Ford Nuclear Reactor (FNR), Massachusetts Institute of Technology Reactor (MITR), Georgia Institute of Technology Research Reactor (GTRR), and University of Wisconsin ...


Simulation Of Plenum Thermo-Hydraulics In A Liquid Metal Fast Breeder Reactor Under A Buoyancy-Affected Condition , Min-Jen Chen Jan 1983

Simulation Of Plenum Thermo-Hydraulics In A Liquid Metal Fast Breeder Reactor Under A Buoyancy-Affected Condition , Min-Jen Chen

Retrospective Theses and Dissertations

The thermohydraulics of the LMFBR upper plenum under buoyancy - affected condition has been investigated. It was found that under low flow, thermal stratified conditions, reduced-scale sodium simulation cannot accurately predict the full-scale LMFBR upper plenum behavior. However, water is an adequate test fluid for prediction of full-scale LMFBR thermohydraulic behavior. It was also found that the threshold Peclet number above which convection dominates the heat transfer is about 10;A three-dimensional transient computer code PLENMIX which is a simplified version of COMMIX-1A has been used for the analysis. It was found that the PLENMIX code can accurately predict temperature transients ...