Nuclear Engineering Commons^™

Hybrid Machine Learning And Comparison Error Minimization For Frequency Domain-Based Rapid State Estimation In Structures Subjected To High-Rate Boundary Change, James Scheppegrell

Theses and Dissertations

Dynamic forces and evolving structural boundary conditions pose challenges for various structural systems such as aircraft, orbital infrastructure, and energy harvesting devices. The design, evaluation, and functionality, of such systems can be aided through the collection and analysis of data. However, real-time decision-making for systems experiencing high-rate changes can pose unique challenges, if assessments are to be made accurately and rapidly enough to be relevant. In cases where the systems are well-defined and thoroughly understood, monitoring the frequency response can be instrumental in determining the state of structures subjected to high-rate structural boundary condition changes. This study focuses on investigating …

Assessing The Feasibility Of Transportable Nuclear Reactors: A Radiological Risk Case Study, Robert John Demuth

Theses and Dissertations

The feasibility of a truly transportable nuclear reactor presents significant radiological challenges that must be addressed to ensure compliance with regulatory limits and operational safety. This study evaluates the radiological constraints associated with the transportability of the Transportable Helium-cooled One-megawatt Reactor (THOR), focusing on neutron activation, shielding performance, and external radiation exposure. Using advanced radiation transport and activation modeling within the SCALE framework, a comprehensive analysis was conducted to quantify dose rates during operation, post-shutdown, and transport. A key aspect of this work involved assessing activation products in core materials, auxiliary components, and surrounding environments to determine their contribution to …

Conceptual Design And Preliminary Safety Analysis Of A Proposed Nuclear Microreactor For Mobile Application, A. S. M. Fakhrul Islam

Theses and Dissertations

Conceptual design and preliminary safety analysis of a proposed helium-cooled Micro Nuclear reactor In ONe megawatt (MINION) thermal output is done through three-dimensional computational analysis. Compared with the conventional high temperature gas-cooled reactors (HTGR) that use tri-structural or bi-structural isotropic (TRISO or BISO) fuel particles, MINION uses traditional clad fuel pellets. This greatly simplifies the core design and allows for a substantially more compact design thanks to the fuel’s signif-icantly higher U-235 atom density. Uranium carbide (UC) enriched up to 19.75% in U-235 is used as fuel. In the power-flattened design, U-235 enrichment varies radially between 19.75% and 10.5% in …

Thermodynamic Assessment Of Chromium Corrosion In The Na-K-Mg-U(Iii, Iv) Chloride Salt, Jacob Allen Yingling

Theses and Dissertations

Thermodynamic descriptions of high-order molten chloride salt systems are essential for the development of fast molten salt reactor (MSR) technologies. However, a complete thermodynamic assessment of the essential Na-K-Mg-U(III)-U(IV) molten chloride salt with the most prevalent CrCl2 corrosion product has yet to be provided. This is remedied in the present work through application of the CALculation of PHAse Diagrams (CALPHAD) approach to the available thermodynamic data and new measurements that include differential scanning calorimetry (DSC) observations for previously unexplored two and three component chloride salt systems. Through these efforts, a unique approach was developed for the quantification of uncertainty in …

Nuclear Fuel Cycle: Safe Management Of Spent Nuclear Fuel, Robert L. Sindelar

Journal of the South Carolina Academy of Science

The aim for storage of spent nuclear fuel (SNF) either in wet or in dry storage systems is to ensure general safety objectives are met throughout a desired storage period. Staff at the Savannah River National Laboratory (SRNL), in collaborations with partners at other national laboratories, industry research organizations, and the University of South Carolina (UofSC), have performed materials aging testing and analyses, and have established nuclear materials aging management programs to support extended periods of safe storage of research reactor (RR) SNF and of commercial power reactor (PR) SNF pending ultimate disposal. Several example challenges include susceptibility of aluminum …

Thermal-Hydraulic System Analysis Of A Proposed 1 Mwth Nuclear Gas Cooled Microreactor, Aaron S. Fernandez

Theses and Dissertations

Nuclear power plants have historically been large, expensive, and prolonged construction projects. A new generation of reactors categorized as microreactors is currently being designed to address energy needs in remote locations. The University of South Carolina has proposed the MIcro Nuclear In ONe megawatt (MINION-1) design to achieve ten years of operation. The primary loop transfers 1 MW of energy from the reactor core using helium, while the secondary uses air in an open Brayton cycle to produce power and heat. This work focuses on modeling the MINION-1 design using the System Analysis Module (SAM) developed by Argonne National Labs …

Mechanistic Multiphysics Modeling Of Cladding Rupture In Nuclear Fuel Rods During Loss-Of-Coolant Accident Conditions, Kyle Allan Lawrence Gamble

Theses and Dissertations

The Loss of Coolant Accident (LOCA) is a design basis accident that is included as part of the safety analysis of nuclear power plants. As the nuclear industry desires to increase the cycle length and discharge burnup of existing nuclear power plants they must demonstrate safe operation during a LOCA on high burnup fuel. During a LOCA transient on high burnup fuel rods, the rods may undergo a process known as fuel fragmentation, relocation, and dispersal (FFRD). To permit dispersal, the cladding encapsulating the fuel must rupture with an opening size large enough to allow the fragmented fuel particles to …

Experimental Evaluation Of Drying Spent Nuclear Fuel For Dry Cask Storage Through Vacuum And Forced Helium Dehydration, Jonathan Ellis Perry

Theses and Dissertations

Currently approximately 80,000 tons of spent nuclear fuel (SNF) are in either dry or wet storage at nuclear power plants or designated spent fuel storage facilities. Over 2,000 tons are added to that sum each year. All SNF must be dried to an accepted Nuclear Regulatory Commission (NRC) drying criteria before being placed in dry storage. Inadequately drying SNF increases the risk of detrimental effects to the fuel while in dry storage. Two commonly used drying operations in the nuclear industry are vacuum drying and forced helium dehydration (FHD). Although these drying processes have been used for many years, there …

Computational Modeling Of Radiation Damage In A Multi-Phase Ceramic Waste Form Using Moose, Zeyu Chen

Theses and Dissertations

Ceramic waste forms have been proposed to replace the traditional glassy waste forms for long term stabilization of radionuclides. These waste forms are constantly exposed to self-irradiation emitted from the constituent radionuclides causing their material properties to change accordingly. It has been known that the radiation damage in waste forms is dominated by alpha particles emitted from transuranic (TRU) radionuclides. Since alpha particles usually have a range of 10~20 μm in such waste forms, some fraction of any non-transuranic containing phases (for a multiphase waste form) will be undamaged (or less damaged) if containing large enough grain sizes. Modeling and …

Characterization And Drying Of Oxyhydroxides On Aluminum Clad Spent Nuclear Fuel, Matthew Shalloo

Theses and Dissertations

Research reactors such as the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR) employ aluminum-clad fuel elements made up of many thin plates with uranium dispersed within. In most engineering applications, aluminum is considered to have favorable corrosion characteristics. It forms a thin oxide layer [Al2O3] under atmospheric conditions that is impenetrable to oxygen thus stopping any further corrosion. However, both aluminum metal and Al2O3 react with water to form hydrous oxides which are less protective against further corrosion and form significantly thicker layers than oxidation in dry air. As a result, aluminum-clad spent nuclear fuel (ASNF) …

Modeling Neutron Interaction Inside A 2d Reactor Using Monte Carlo Method, A. S. M. Fakhrul Islam

Theses and Dissertations

Scientists and engineers have been working for many years to develop accurate approaches to analyzing nuclear power reactors using computer codes that closely model the behavior of neutrons in a reactor core. The Monte Carlo simulation method is capable of treating complex geometries with a high level of resolution and fidelity to model neutron interactions inside a reactor core. With the requirement of accurate modeling in reactor physics and dynamics and great innovation of computer technology, Monte Carlo method is becoming an ever more powerful tool and receiving rising attention. In this study, Monte Carlo method is used to model …

Bison Simulation-Based Identification Of Important Design Criteria For U3si2 Fuels With Composite-Monolithic Duplex Sic Cladding, Jacob A. Yingling

Theses and Dissertations

Accident Tolerant Fuels (ATF) require a combination of fuel and cladding which have comparable longevity characteristics to UO2 while improving resistance to radiological release during and after accidents. U3Si2 has been proposed for use in ATF concepts for its high uranium density and high thermal conductivity which provide improved fuel performance. However, some of U3Si2’s material properties are not well understood. One such property, thermal creep of U3Si2, is an important contributing factor to U3Si2’s viability as an ATF. No experimentally derived thermal creep model is published for U3Si2, and previous analyses of compressive thermal creep experimental data lack statistical …

Modeling The Uranium-Silicon Phase Equilibria Based On Computational And Experimental Analysis, Tashiema Lixona Ulrich

Theses and Dissertations

As part of Accident tolerant fuel initiative, the uranium-silicide compound, U3Si2, is under consideration as a potential replacement for conventional uranium dioxide fuel. It is of interest as its higher uranium density of 11.3 g(U)/cm3 compared to 9.7 g(U)/cm3 for UO2 may allow use of more robust, but less neutronically economical fuel cladding. The improved uranium content would not only accommodate the neutronic penalty inherent to certain accident tolerant cladding concepts but also facilitate improved reactor performance with the potential for longer fuel cycles.

The U-Si system has been the subject of various studies that mainly focused on thermophysical properties, …

Implementation Of View Factor Model And Radiative Heat Transfer Model In Moose, Abdurrahman Ozturk

Theses and Dissertations

View factors are functions that represent the geometric relationship between surfaces. They are important parameters for radiative heat transfer calculations. View factor catalogues are available for simple geometries in the current literature. However, in the case of complicated geometry, analytical or numerical methods are needed to evaluate view factors. The Monte Carlo (MC) method is the most flexible one among numerical methods, which are used to calculate view factors, since it can be applied to any geometry.

When experimental studies are not affordable to conduct, modeling of engineering problems gains more importance. Idaho National Laboratory (INL)’s finite element framework Multiphysics …

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

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

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, …

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 …

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 …

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 …

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 possible …

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 be …

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 …

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 …

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

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 …

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 with …

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 evaluation of …

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 …

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 …