An overall systems design code was developed to model an advanced in-core thermionic energy conversion based nuclear reactor system for space applications at power levels of 10 to 50 kWe. The purpose of this work was to provide the overall shell for the systems code and to also provide the detailed neutronic analysis section of the code. The design code that was developed is to be used to evaluate a reactor system based upon a single cell thermionic fuel element which uses advanced technology to enhance the performance of single cell thermionic fuel elements. A literature survey provided information concerning how other organizations performed system studies on similar space reactor designs. / Graduation date: 1994

Incore thermionic space reactor design concepts which operate at a nominal power output range of 20 to 50 kWe are described. Details of the neutronic, thermionic, thermal hydraulics and shielding performance are presented. Due to the strong absorption of thermal neutrons by natural tungsten, and the large amount of that material within the reactor core, two designs are considered. An overall system design code has been developed at Oregon State University to model advanced incore thermionic energy conversion based nuclear reactor systems for space applications. The code modules include neutronics and core criticality, a thermionic fuel element performance module with integral thermal hydraulics calculation capability, a radiation shielding module, and a module for the waste heat rejection. The results show that the driverless single cell ATI configuration, which does not have driver rods, proved to be more efficient than the driven core, which has driver rods. It also shows that the inclusion of the true axial and radial power distribution decrease the overall conversion efficiency. The flattening of the radial power distribution by three different methods would lead to a higher efficiency. The results show that only one thermionic fuel element (TFE) works at the optimum emitter temperature; all other TFEs are off the optimum performance and result in 40 % decrease of the efficiency of the overall system. / Graduation date: 1993

Bibliography: p. 253-257.

Thesis (Ph. D.)--University of Missouri-Columbia, 2006. / The entire dissertation/thesis text is included in the research.pdf file; the official abstract appears in the short.pdf file (which also appears in the research.pdf); a non-technical general description, or public abstract, appears in the public.pdf file. Title from title screen of research.pdf file viewed on (March 2, 2007) Vita. Includes bibliographical references.

A large set of experimental data, obtained in Russia, was analyzed and a new heat-transfer correlation for supercritical water was developed. This experimental dataset was obtained within conditions similar to those for proposed SuperCritical Water-cooled nuclear Reactor (SCWR) concepts. Thus, this new correlation, for forced convective heat transfer in the normal heat-transfer regime, can be used for preliminary heat-transfer calculations in SCWR fuel channels. It has demonstrated a good fit for Heat Transfer Coefficient (HTC) values (±25%) and for wall temperature calculations (±15) for the analyzed dataset. This correlation can be used for supercritical water heat exchangers linked to indirectcycle concepts and the co-generation of hydrogen, for future comparisons with other independent datasets, with bundle data, as the reference case, for the verification of computer codes for SCWR core thermalhydraulics and for the verification of scaling parameters between water and modeling fluids. / UOIT

Preliminary nuclear design studies have been completed on ten configurations of nuclear reactors for low power (1-30 kWe) space applications utilizing thermionic energy conversion. Additional design studies have been conducted on the TRICE multimegawatt in-core thermionic reactor configuration. In each of the cases, a reactor configuration has been determined which has the potential for operating 7 years with sufficient reactivity margin. Additional safety evaluations have been conducted on these configurations including the determination of sufficient shutdown reactivity, and consideration of water immersion, water flooding, sand burial, and reactor compaction accident scenarios. It has been found, within the analysis conducted using the MCNP Monte Carlo neutron transport code, that there are configurations which are feasible and deserve further analysis. It has also been found that solid core reactors which rely solely on conduction for heat removal as well as pin type cores immersed in a liquid metal bath have merit. The solid cores look attractive when flooding and compaction accident scenarios are considered as there is little chance for water to enter the core and cause significant neutron moderation. A fuel volume fraction effect has also been found in the consideration of the sand burial cases for the SP-100 derived configurations. / Graduation date: 1994

Several computer codes based on one and two-group diffusion theory models were developed for SHUFFLE. The programs were developed to calculate power distributions in a two-dimensional quarter core geometry of a pressurized power reactor. The various coarse-mesh numerical computations for the power calculations yield the following: the Borresen's scheme applied to the modified one-group power calculation came up with an improved power distribution, the modified Borresen's method yielded a more accurate power calculations than the Borresen's scheme, the face dependent discontinuity factor method have a better prediction of the power distribution than the node averaged discontinuity factor method, Both the face dependent discontinuity factor method and the modified Borresen's methods for the two-group model have quite attractive features. / Graduation date: 1992

This thesis describes the development of computer programs and the theory for the calculations of stopping power and nuclear reaction kinematics. The nuclear reaction kinematics program computes position and nonrelativistic energy data as well as center-of-mass solid angle transformations and information on detector resolution for nuclear reactions and elastic scattering experiments involving two-body final states. The stopping power program calculates stopping power (an index of the charged particle energy absorption properties of a material) of elemental absorbers for protons, deuterons, tritons, He3, and alpha particles from minimal input data. The calculated stopping powers are accurate to within one per cent over the nonrelativistic energy range of 2 to 12 Mev.Ball State UniversityMuncie, IN 47306

The solution of coupled neutronic/thermal hydraulic nuclear reactor calculations requires the treatment of the nonlinear feedback induced by the thermal hydraulic dependence of the neutron cross sections. As a result of these nonlinearities, current solution techniques often diverge during the iteration process. These instabilities arise due to the low level of coupling achieved by these methods between the neutronic and thermal hydraulic components. In this work, this solution method is labeled the Decoupled Iteration (DI) method, and this technique is examined in an effort to improve its efficiency and stability. An examination of the DI method also serves to provide insight into the development of more highly coupled iteration methods. After the examination of several possible iteration procedures, two techniques are developed which achieve both a higher degree of coupling and stability. One such procedure is the Outer Iteration Coupling (OIC) method, which combines the outer iteration of the multigroup diffusion calculation with the controlling iteration of the thermal hydraulic calculations. The OIC method appears to be stable for all cases, while maintaining a high level of efficiency. Another iteration procedure developed is the Modified Axial Coupling (MAC) procedure, which couples the neutronic and thermal hydraulic components at the level of the axial position within the coolant channel. While the MAC method does achieve the highest level of coupling and stability, the efficiency of this technique is less than that of the other methods examined. Several characteristics of these coupled calculation methods are examined during the investigation. All methods are shown to be relatively insensitive to thermal hydraulic operating conditions, while the dependence upon convergence criteria is quite significant. It is demonstrated that the DI method does not converge for arbitrarily small convergence criteria, which is a result of a non-asymptotic solution approximation by the DI method. This asymptotic quality is achieved in the coupled methods. Thus, not only do the OIC and MAC techniques converge for small values of the relevant convergence criteria, but the computational expense of these methods is a predictable function of these criteria. The degree of stability of the iterative techniques is enhanced by a higher level of coupling, but the efficiency of these methods tends to decrease as a higher degree of coupling is achieved. This is apparent in the diminished efficiency of the MAC procedure. Seeking an optimum balance of efficiency and stability, the OIC technique is demonstrated to be the optimum method for coupled neutronic/thermal hydraulic reactor calculations. / Graduation date: 1994

In a nuclear power plant, one of the most important systems for both safety and performance is the reactor cooling system. The cooling system is generally driven by one or more very large centrifugal pumps. Most reactor coolant pumps utilize a multi-stage mechanical face seal system for fluid containment. As a result, these seal systems are critical to safe, continued operation of a nuclear reactor. Without adequate sealing, loss of coolant volume can occur, and a reactor may be forced to shut down, costing the operating utility significantly until it can be brought online again. The main advantage of mechanical face seals is their self-adjusting properties. These seals are tuned so that they automatically adjust to varying fluid conditions to provide adequate leakage control. Because of the enormous pressures, complicated water chemistry, and possible large temperature transients, the mechanical seals inside a reactor coolant pump must be some of the most robust seals available. In addition, their long service life and continuous operation demand durability and the capability to adjust to a wide range of conditions. However, over time, wear, chemical deposition, or changing operating conditions can alter the face gap, which is the critical geometry between the sealing faces of a seal. An altered face gap can lead to undesirable conditions of too much or not enough leakage, which must be maintained within a certain range to provide lubrication and cooling to the seal faces without resulting in uncontrolled coolant volume loss. Nuclear power plants operate within strict leakage ranges, and long-term effects causing undesirable leakage can eventually necessitate a reactor shutdown if the seal cannot self-adjust to control the leakage. This document will examine possible causes of undesirable leakage rates in a commonly-used reactor coolant pump assembly. These causes will be examined to determine the conditions which promote them, the physical explanation for their effect on the operation of a mechanical seal, and possible methods of mitigation of both the cause and its effect. These findings are based on previous publications by utilities and technical and incident reports from reactor stations which detail actual incidents of abnormal seal performance and their root causes as determined by the utilities. Next, a method of increasing the ability of a mechanical seal to adapt to a wider range of conditions will be proposed. This method involves modifying an existing seal face to include a method of active control. This active control focuses on deliberately deforming one face of the mechanical sealing face pair. This deformation alters the face gap in order to make the fluid conditions inside the face gap more preferable, generating more or less leakage as desired. Two methods of actuation, hydraulic pressure and piezoelectric deformation, will be proposed. Finally, a model of the actively controlled seal faces will be introduced. This model includes a method of numerically solving the Reynolds equation to determine the fluid mechanics that drive the lubrication problem in the seal face and coupling the solution with a deformation analysis in a finite element model of a seal face. The model solves iteratively until a converged solution of a sealed pressure distribution, a resulting face deformation, and a calculated leakage rate is reached. The model includes a study of the effects of induced deformation in the seal via both hydraulic and piezoelectric actuation and the ability of this deformation to control the leakage rate.