Master of Science / Department of Mechanical and Nuclear Engineering / Akira T. Tokuhiro / The Secure Transportable Autonomous Reactor – Liquid Metal system combines a Generation IV nuclear reactor with an advanced Supercritical Carbon Dioxide (S-CO[subscript]2) Brayton power conversion cycle. The Brayton cycle was selected as the power conversion cycle due to its high efficiency, small turbomachinery size, and competitive cost due to reduced complexity as compared to a traditional Rankine cycle. Overall system thermal efficiency is closely tied to the performance of the precooler and recuperators. The Printed Circuit Heat Exchanger (PCHE) manufactured by Heatric is being considered for use as both the precooler and recuperator in the STAR-LM system due to its high effectiveness, wide temperature and pressure operating range, small size, and low cost. PCHEs have been used primarily in the hydrocarbon processing industry to date, and are relatively new in being considered for nuclear applications. In this study, a PCHE is investigated using S-CO[subscript]2 and water as the heat transfer media in conditions relevant to the precooler in the STAR-LM system. Experiments conducted with small temperature differences across the PCHE revealed that the heat transfer coefficient is strongly correlated with the temperature-dependent specific heat near the pseudocritical point. The STAR-LM precooler outlet temperature is near the pseudocritical point, making this region of interest to this work. Testing was conducted to determine the effect of property variation near the precooler outlet in conditions with large temperature differences in the PCHE. These tests revealed that maintaining the precooler outlet temperature near the pseudocritical point does not have a significant effect on heat transfer coefficients in the PCHE under large temperature difference test conditions. Computational Fluid Dynamics (CFD) models were developed to simulate fluid flow and heat transfer in the PCHE. A 2D, 4-channel, zig-zag model was found to reproduce the outlet temperatures to within approximately 15% relative error. The 3D straight channel model reproduced the experimental data to within 3% relative error for the cases simulated. Both of these models predicted the water side outlet temperatures to within 20% relative error.

Master of Science / Department of Architectural Engineering and Construction Science / Kimberly W. Kramer / This report explores the requirements for the design and analysis of concrete containment and shows how newer material technologies such as self-consolidating concrete (SCC) and fiber reinforcement could assist in the constructability and durability of new nuclear power plant facilities. SCC for example, enables concrete to flow in the forms around the reinforcement and provides a more uniform adhesion with the reinforcement. Additionally, fiber reinforcement in the concrete mix increases bonding capability, thus making the concrete less likely to fracture. In particular, the ease of constructability benefits offshore floating nuclear power plants and preapproved modular power plants. To differentiate, the offshore plant would employ the assembly line to make all the plants the same while the modular plant, designed to be used anywhere, is not site specific and is typically smaller. Regarding research method, the report starts with the history of the nuclear industry in the United States, including the last nuclear power plant constructed, clarifying that nuclear energy was first harnessed for a submarine propulsion system before being employed to generate electricity. After these early endeavors, two major accidents, Three Mile Island (March 28, 1979) and Chernobyl (April 26, 1986), provided information regarding the lack of safety of nuclear power plant design and operation. Since the containment building is the focus of this report, recognizing the loads and the load combinations for design was the next step in research. Following that, the next step was to determine the design considerations and analyze the containment structure. New material technologies clearly have opened the door to new construction techniques, and the combination of new materials and methods offers structural engineers opportunity to build inherently safer nuclear power plants.

Doctor of Philosophy / Department of Mechanical and Nuclear Engineering / Douglas S. McGregor / Cadmium Zinc Telluride (CdZnTe) has been used for many applications, such as medical imaging and astrophysics, since its first demonstration as a room temperature operating gamma-ray detector in 1992. The wide band gap, high effective Z-number and high resistivity of CdZnTe make it a good candidate for use as a room temperature operated detector with good absorption efficiency, while maintaining a low bulk leakage current at high electric fields. Nevertheless, the low mobility lifetime products mu tau of holes in CdZnTe makes detectors position sensitive, unless advanced detector designs are employed. Among those designs is the Frisch collar technology which turns the detector into a single carrier device by negating the degrading effects of hole trapping and low mobility. The superiority of the Frisch collar technology over other methods include its inexpensive associated electronics and straight forward fabrication process. The main objective of this research study is to develop a large volume gamma-ray detector with an array of individual CdZnTe Frisch collar gamma-ray spectrometers while still using a single readout. Several goals were to be accomplished prior to the main objective. One goal is to develop a reliable low cost method to fabricate bulk CdZnTe crystals into Frisch collar detectors. Another goal was to investigate the limitations of crystal geometry and the crystal electrical properties to obtain the best spectroscopic performance from CdZnTe Frisch collar detectors. Still another goal was to study all other external parameters such as the collar length, anode to cathode ratio, the insulator thickness and applied voltage on performance of CdZnTe Frisch collar detectors. The final goal was to construct the CdZnTe Frisch collar devices into an array and to show its feasibility of being used for large volume detector.

Master of Science / Department of Mechanical and Nuclear Engineering / J. Kenneth Shultis / A program which calculates the radiation dosage to a predetermined set of components inside of a kill vehicle as a result of natural or artificial radiation sources has been developed for use within the confines of a parent external simulation. This dose can then be used to determine if a critical component has malfunctioned or failed completely, thereby rendering the interceptor unable to finish its mission. Knowledge of system and component performance as a function of incident high-energy particles leads to better battle management planning, CONOPS, and potentially a more efficient shielding design to achieve a higher probability of mission success.

Master of Science / Department of Mechanical and Nuclear Engineering / J. Kenneth Shultis / Perforated neutron detectors suffer the unfortunate effect that their efficiency is a strong function of the direction of neutron incidence. It is found, by Monte Carlo simulation of many perforation shapes, that sinusoidal-type perforations greatly reduce the variation of detector efficiency. Detectors with rod-type perforations are modeled using a hybrid transport method linking the MCNP transport code and a specialized ion-transport code to calculate the probability that a neutron is detected. Channel, chevron, and sinusoidal perforations are modeled using other customized transport codes. Detector efficiency calculations are performed for neutrons incident at various polar and azimuthal angles. It is discovered that the efficiency losses of the detectors result from the decreasing solid angle subtended by the detector from the source and streaming through the detector at specific azimuthal angles. Detectors achieving an efficiency in excess of 10% and having a relatively flat ± 1% angular dependence in all azimuthal angles and polar angles between 0 and 60 degrees are predicted. Efficiencies up to 25% are achievable at the loss of directional independence. In addition to minimizing the directional dependence of the perforated detectors, the feasibility of developing a neutron detector for deployment in cargo containers to locate nuclear weapon pits is investigated using the MCNP transport code. The detector considered is a 7-mm diameter, 6LiF, rod-perforated detector surrounded in a cylinder of polyethylene. The optimum thicknesses of surrounding polyethylene, to maximize the response of the detector, is determined to be 10 cm of radial, 5 cm of front, and 5 cm of back polyethylene for end-on neutron incidence. Such a detector is predicted to produce a count rate between 12 and 15 cpm from a nuclear-weapon pit composed of 90% 239Pu and 10% 240Pu at a distance of 3 m. Side incidence is also considered, and the optimum moderator dimensions are 8 cm of radial, 10 cm of front, and 10 cm of back polyethylene that produce approximately the same count rate.

Master of Science / Department of Mechanical and Nuclear Engineering / Donald L. Fenton / Increase in demand for energy necessitates nuclear power units to increase their peak power limits. This increase implies significant changes in the design of the nuclear power unit core in order to provide better economy and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Water Reactors (PWR) is the so called ‘Axial Offset Anomaly (AOA)’. An Axial Offset Anomaly (AOA) is the unexpected change in the core axial power distribution during the operation of a PWR from the predicted distribution. This problem is thought to be occurring because of precipitation and deposition of lithiated compounds such as lithium metaborate (LiBO[subscript]2) on the fuel rod. Due to its intrinsic property, the deposited boron absorbs neutrons thereby affecting the total power distribution in the reactor. AOA is thought to occur when there is sufficient build up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult because there is little information regarding the heat and mass transfer during subcooled nucleate boiling. This thesis describes the experimental investigation that was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of LiBO[subscript]2 and boric acid (H[subscript]2BO[subscript]3) solutions along with deionized water. The experimental data collected includes the effect of coolant concentration, degree of subcooling, system pressure and heat flux on pool boiling heat transfer coefficients. An analysis of deposits formed on the fuel rod during subcooled nucleate boiling is also included in the thesis. The experimental results reveal that the pool boiling heat transfer coefficient is degraded by the presence of boric acid and lithium metaborate in water. At concentration of 5000 ppm in water, the boric acid solution reduced the heat transfer coefficient by 23% and lithium metaborate solution reduced the heat transfer coefficient by 26%.

Master of Science / Department of Mechanical and Nuclear Engineering / Douglas S. McGregor / Purification of cadmium, zinc and tellurium was attempted to improve the quality of cadmium-zinc-telluride (CdZnTe) crystal growth. Specifically, vacuum distillation, zone refining and H[subscript]2 gas flow assisted zone refining were all investigated as methods to purify the constituent elements of CdZnTe. A unique multi-chamber ampoule was used to enable a purification sequence starting with double vacuum distillation followed by zone refining all without sample handling after the initial step. Modifications due to unique material properties of Cd and Zn were developed. Glow discharge mass spectroscopy (GDMS) analysis was used to measure impurity concentrations of 74 elements. Cd purification using vacuum distillation proved to be an effective method to reduce the impurity level of 5N starting material to a purity between the range of 6N5 and 7N5, as measured using GDMS and laser ablation mass spectroscopy. Combined Zn double vacuum distillation and zone refining in an enclosed Ar atmosphere using 5N starting material yielded material with a purity between the range of 5N8 to 6N8. Tellurium purification using combined double vacuum distillation followed by zone refining under continuous H[subscript]2 flow of 4N specified raw material resulted in high purity tellurium between the range of 6N3 and 7N4.

Master of Science / Department of Mechanical and Nuclear Engineering / William L. Dunn / A prototype system for detecting explosives at standoff distances, using a signature based radiation scanning approach, is being developed at Kansas State University. The prototype will incorporate both a machine x-ray source and a machine neutron source to generate signatures from unknown samples of material. These signatures can be compared to templates measured or calculated from known explosive samples using a figure-of-merit. The machine neutron source uses the fusion of deuterium and tritium to create 14.1 MeV neutrons. Due to its radioactivity, the tritium must be sealed within the system. A new method of controlling the gas pressure with the DT generator was developed using a Zr-V-Fe getter supplied by a commercial firm. The shielding and collimation of the 14.1 MeV neutron source is accomplished using layers of steel, high-density polyethylene and borated high-density polyethylene. This thesis describes the development of the gas control method for the sealed neutron source, design studies for the shielding and collimation of the neutron source and modifications made to the building in which the prototype is being housed.