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Full-Text Articles in Nuclear Engineering

Modeling Complex Oxides: Thermochemical Behavior Of Nepheline-Forming Na-Al-Si-B-K-Li-Ca-Mg-Fe-O And Hollandite-Forming Ba-Cs-Ti-Cr-Al-Fe- Ga-O Systems, Stephen A. Utlak Apr 2019

Modeling Complex Oxides: Thermochemical Behavior Of Nepheline-Forming Na-Al-Si-B-K-Li-Ca-Mg-Fe-O And Hollandite-Forming Ba-Cs-Ti-Cr-Al-Fe- Ga-O Systems, Stephen A. Utlak

Theses and Dissertations

High concentrations of Na2O and Al2O3 in the liquid high-level radioactive waste (HLW) stored at the Hanford Site can cause nepheline (NaAlSiO4) to precipitate in a vitrified monolithic waste form upon cooling. Nepheline phase formation removes glass- former SiO2 and -modifier Al2O3 from the immobilization matrix in greater proportion to alkalis, which can reduce glass durability and consequently increase the leach rate of radionuclides into the surrounding environment.

Current uncertainty in defining the HLW glass composition region prone to precipitating nepheline necessitates targeting a conservative waste loading, which raises operational costs by extending the liquid radioactive waste disposal mission and ...


Modeling Complex Oxides: Thermochemical Behavior Of Nepheline-Forming Na-Al-Si-B-K-Li-Ca-Mg-Fe-O And Hollandite-Forming Ba-Cs-Ti-Cr-Al-Fe- Ga-O Systems, Stephen A. Utlak Apr 2019

Modeling Complex Oxides: Thermochemical Behavior Of Nepheline-Forming Na-Al-Si-B-K-Li-Ca-Mg-Fe-O And Hollandite-Forming Ba-Cs-Ti-Cr-Al-Fe- Ga-O Systems, Stephen A. Utlak

Theses and Dissertations

High concentrations of Na2O and Al2O3 in the liquid high-level radioactive waste (HLW) stored at the Hanford Site can cause nepheline (NaAlSiO4) to precipitate in a vitrified monolithic waste form upon cooling. Nepheline phase formation removes glass- former SiO2 and -modifier Al2O3 from the immobilization matrix in greater proportion to alkalis, which can reduce glass durability and consequently increase the leach rate of radionuclides into the surrounding environment.

Current uncertainty in defining the HLW glass composition region prone to precipitating nepheline necessitates targeting a conservative waste loading, which raises operational costs by extending the liquid radioactive waste disposal mission and ...


Mechanical Characterization And Non-Destructive Evaluation Of Sicf-Sicm Composite Tubing With The Impulse Excitation Technique, Nathaniel Truesdale Jan 2017

Mechanical Characterization And Non-Destructive Evaluation Of Sicf-Sicm Composite Tubing With The Impulse Excitation Technique, Nathaniel Truesdale

Theses and Dissertations

With growing interest on ceramic fiber reinforced ceramic matrix composites (CMC) for accident tolerant fuel, the need for mechanical characterization of ceramic composite arises. It has been of particular interest to non-destructively evaluate the mechanical performance of these composites. Impulse excitation (IE) is a well-established method for non-destructive mechanical characterization of homogeneous isotropic material of well-defined shapes. In this thesis, impulse excitation technique was applied for non-destructive characterization of composite tube for the first time as far as we know. CMC, when stressed beyond its damage threshold, will experience various forms of structural damage, such as matrix micro-cracking, fiber-matrix debonding ...


Deformation Induced Martensitic Transformation In 304 Stainless Steels, Junliang Liu Jan 2016

Deformation Induced Martensitic Transformation In 304 Stainless Steels, Junliang Liu

Theses and Dissertations

304 stainless steel is an austenitic steel widely used for various applications due to a good combination of strength and ductility and relative low cost. It is known to be metastable as the austenite phase can transform into martensite under stress. In this work, a new method (in-situ tensile TEM) and the traditional method (ex-situ tensile tests and TEM, XRD characterization) were used to investigate the mechanisms of deformation-induced martensitic transformation in 304SS samples at different temperatures.

The ex-situ tensile tests were conducted under a strain rate of 10-3 s-1 until rupture. After the tensile tests, the fractured area was ...


Dosimetry, Activation, And Robotic Instrumentation Damage Modeling Of The Holtec Hi-Storm 100 Spent Nuclear Fuel System, C. Ryan Priest Jan 2016

Dosimetry, Activation, And Robotic Instrumentation Damage Modeling Of The Holtec Hi-Storm 100 Spent Nuclear Fuel System, C. Ryan Priest

Theses and Dissertations

The Holtec HI-STORM 100 spent fuel storage system is an intermediate storage mechanism for SNF assemblies. Long term licensing and storage requires consideration of material degradation of the stainless steel multi-purpose canister holding the spent fuel. This material degradation is predicted to take the form of environmentally assisted cracking on weld interfaces of the stainless steel cylinder. Several of these spent fuel storage arrays are located in coastal environments. These sites include both the Hope Creek and Diablo Canyon nuclear power stations. A multi-sensory robotic package is being designed for non-destructive assay of the cask annular environment. It will be ...


Analysis Of Pellet Cladding Interaction And Creep Of U3si2 Fuel For Use In Light Water Reactors, Kathryn E. Metzger Jan 2016

Analysis Of Pellet Cladding Interaction And Creep Of U3si2 Fuel For Use In Light Water Reactors, Kathryn E. Metzger

Theses and Dissertations

Following the accident at the Fukushima plant, enhancing the accident tolerance of the light water reactor (LWR) fleet became a topic of serious discussion. Under the direction of congress, the DOE office of Nuclear Energy added accident tolerant fuel development as a primary component to the existing Advanced Fuels Program. The DOE defines accident tolerant fuels as fuels that “in comparison with the standard UO2- Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or ...


Intercode Advanced Fuels And Cladding Comparison Using Bison, Frapcon, And Femaxi Fuel Performance Codes, Aaren Rice Dec 2015

Intercode Advanced Fuels And Cladding Comparison Using Bison, Frapcon, And Femaxi Fuel Performance Codes, Aaren Rice

Theses and Dissertations

The high density uranium-based fuels are regaining popularity as the current fleet of LWR’s are showing interest in uprating plants to increase accident tolerance and performance. Fuels such as U3Si2, UN, and UC all contain a higher uranium loading and thermal conductivity than that of UO2 making them attractive in combination with an advanced cladding type such as the ceramic SiC cladding. In addition to adding more mass uranium to the core without surpassing current enrichment limits, these advanced fuels and claddings are designed with increased accident tolerance performance in a LOCA type scenario in mind. One of the ...


Implementation And Evaluation Of Fuel Creep Using Advanced Light-Water Reactor Materials In Frapcon 3.5, Spencer Carroll Jan 2014

Implementation And Evaluation Of Fuel Creep Using Advanced Light-Water Reactor Materials In Frapcon 3.5, Spencer Carroll

Theses and Dissertations

As current reactors approach the end of their operable lifetime, new reactors are needed if nuclear power is to continue being generated in the United States. Some utilities have already began construction on newer, more advanced LWR reactors, which use the same fuel as current reactors and have a similar but updated design. Others are researching next generation (GEN-IV) reactors which have new designs that utilize alternative fuel, coolants and other reactor materials. Many of these alternative fuels are capable of achieving higher burnups and are designed to be more accident tolerant than the currently used UO2 fuel. However, before ...


System Analysis With Improved Thermo-Mechanical Fuel Rod Models For Modeling Current And Advanced Lwr Materials In Accident Scenarios, Ian Edward Porter Jan 2014

System Analysis With Improved Thermo-Mechanical Fuel Rod Models For Modeling Current And Advanced Lwr Materials In Accident Scenarios, Ian Edward Porter

Theses and Dissertations

A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will ...


Modified Sodium Diuranate Process For The Recovery Of Uranium From Uranium Hexafluoride Transport Cylinder Wash Solution, Austin Dean Meredith Dec 2013

Modified Sodium Diuranate Process For The Recovery Of Uranium From Uranium Hexafluoride Transport Cylinder Wash Solution, Austin Dean Meredith

Theses and Dissertations

Uranium hexafluoride (UF6) containment cylinders must be emptied and washed every five years in order to undergo recertification, according to ANSI standards. During the emptying of the UF6 from the cylinders, a thin residue, or heel, of UF6 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 ...


The Study Of Alternate, Solid-Phase Fluorinating Agents For Use In Reactive Gas Recycle Of Used Nuclear Fuel, Dillon Inabinett Jan 2013

The Study Of Alternate, Solid-Phase Fluorinating Agents For Use In Reactive Gas Recycle Of Used Nuclear Fuel, Dillon Inabinett

Theses and Dissertations

Surrogate oxides of the Used Nuclear Fuel (UNF) matrix were fluorinated using alternate, solid-phase fluorinating agents XeF2 and NH4HF2 to form volatile and non-volatile compounds and demonstrate the possibility of a chemical and thermal separations. A matrix of experiments was conducted at the milligram quantity scale using a Shimadzu DTG-60 TG/DTA installed at SRNL (Savannah River National Laboratory) for testing of all non-radioactive samples and a Netzsch STA 409 TGA installed in the laboratory at USC (University of South Carolina) for testing of all radioactive samples. The fluorination and subsequent volatilization potentials were analyzed by mixing excess fluorinating agent ...


Characterization Of Two Ods Alloys: 18cr Ods And 9cr Ods, Julianne Kay Goddard Jan 2013

Characterization Of Two Ods Alloys: 18cr Ods And 9cr Ods, Julianne Kay Goddard

Theses and Dissertations

ODS alloys, or oxide dispersion strengthened alloys, are made from elemental or pre-alloyed metal powders mechanically alloyed with oxide powders in a high-energy attributor mill, and then consolidated by either hot isostatic pressing or hot extrusion causing the production of nanometer scale oxide and carbide particles within the alloy matrix; crystalline properties such as creep strength, ductility, corrosion resistance, tensile strength, swelling resistance, and resistance to embrittlement are all observed to be improved by the presence of nanoparticles in the matrix. The presented research uses various methods to observe and characterize the microstructural and microchemical properties of two experimental ODS ...


Advanced Fuels Modeling: Evaluating The Steady-State Performance Of Carbide Fuel In Helium-Cooled Reactors Using Frapcon 3.4, Luke H. Hallman Jan 2013

Advanced Fuels Modeling: Evaluating The Steady-State Performance Of Carbide Fuel In Helium-Cooled 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 gas-cooled 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 ...


Pellet Cladding Mechanical Interactions Of Ceramic Claddings Fuels Under Light Water Reactor Conditions, Bo-Shiuan Li Jan 2013

Pellet Cladding Mechanical Interactions Of Ceramic Claddings Fuels Under Light Water Reactor Conditions, Bo-Shiuan Li

Theses and Dissertations

Ceramic materials such as silicon carbide (SiC) are promising candidate materials for nuclear fuel cladding and are of interest as part of a potential accident tolerant fuel design due to its high temperature strength, dimensional stability under irradiation, corrosion resistance, and lower neutron absorption cross-section. It also offers drastically lower hydrogen generation in loss of coolant accidents such as that experienced at Fukushima. With the implementation of SiC material properties to the fuel performance code, FRAPCON, performances of the SiC-clad fuel are compared with the conventional Zircaloy-clad fuel. Due to negligible creep and high stiffness, SiC-clad fuel allows gap closure ...


Predicting The Crack Response For A Pipe With A Complex Crack, Robert George Lukess Jan 2013

Predicting The Crack Response For A Pipe With A Complex Crack, Robert George Lukess

Theses and Dissertations

Traditional flaw evaluation in the nuclear field uses conservative methods to predict maximum load carrying capacity for flaws in a given pipe. There is a need in the nuclear industry for more accurate estimates of the load carrying capacity of nuclear piping such that probabilistic tools can be used to predict the time to failure for various types of cracks. These more accurate estimates will allow the nuclear industry to repair flaws at a more appropriate time considering external factors such as costs and man-rem planning along with the flaw repair. Analysis of the maximum load carrying capacity of a ...


Evolution Of Microstructure Of Haynes 230 And Inconel 617 Under Mechanical Testing At High Temperatures, Kyle Hrutkay Jan 2013

Evolution Of Microstructure Of Haynes 230 And Inconel 617 Under Mechanical Testing At High Temperatures, Kyle Hrutkay

Theses and Dissertations

Haynes 230 and Inconel 617 are austenitic nickel based superalloys, which are candidate structural materials for next generation high temperature nuclear reactors. High temperature deformation behavior of Haynes 230 and Inconel 617 have been investigated at the microstructural level in order to gain a better understanding of mechanical properties. Tensile tests were performed at strain rates ranging from 10-3-10-5 s-1 at room temperature, 600 °C, 800 °C and 950 °C. Subsequent microstructural analysis, including Scanning Electron Microscopy, Transmission Electron Microscopy, Energy-Dispersive X-ray Spectroscopy, and X-Ray Diffraction were used to relate the microstructural evolution at high temperatures to that of room ...


Fabrication And Characterization Of Surrogate Fuel Particles Using The Spark Erosion Method, Kathryn Elizabeth Metzger Jan 2013

Fabrication And Characterization Of Surrogate Fuel Particles Using The Spark Erosion Method, Kathryn Elizabeth Metzger

Theses and Dissertations

In light of the disaster at the Fukushima Daiichi Nuclear Plant, the Department of Energy's Advanced Fuels Program has shifted its interest from enhanced performance fuels to enhanced accident tolerance fuels. Dispersion fuels possess higher thermal conductivities than traditional light water reactor fuel and as a result, offer improved safety margins. The benefits of a dispersion fuel are due to the presence of the secondary non-fissile phase (matrix), which serves as a barrier to fission products and improves the overall thermal performance of the fuel. However, the presence of a matrix material reduces the fuel volume, which lowers the ...


Application Of Computational Fluid Dynamics Methods To Improve Thermal Hydraulic Code Analysis, Dennis Shannon Sentell, Jr. Jan 2013

Application Of Computational Fluid Dynamics Methods To Improve Thermal Hydraulic Code Analysis, Dennis Shannon Sentell, Jr.

Theses and Dissertations

A computational fluid dynamics code is used to model the primary natural circulation loop of a proposed small modular reactor for comparison to experimental data and best-estimate thermal-hydraulic code results. Recent advances in computational fluid dynamics code modeling capabilities make them attractive alternatives to the current conservative approach of coupled best-estimate thermal hydraulic codes and uncertainty evaluations. The results from a computational fluid dynamics analysis are benchmarked against the experimental test results of a 1:3 length, 1:254 volume, full pressure and full temperature scale small modular reactor during steady-state power operations and during a depressurization transient. A comparative ...