Nuclear Engineering Commons^™

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 …

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 …

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 fissile …

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 …

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 …

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 …

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 …

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 …