lack of consensus among the world researchers on the significance of Po-210 build up in lead cooled reactors. The second objective is to evaluate the advantages and disadvantages of selected candidate metal coolants. In addressing both objectives, the computer code ORIGEN was used. To establish the background basis for these assessments, fundamental concepts of reactor physics are reviewed and discussed.

The behavior of Fe-Cr alloys under irradiation is in part controlled by the characteristics of point defects generated by high energy collision. Radiation enhanced diffusion and radiation induced precipitation are among the mechanisms that lead to changes in the microstructure under irradiation, and are thus controlling effects such as swelling and a' precipitation. Point defects in Fe-Cr alloys are diverse in nature due to their interaction with a variety of local solute configurations. Ab initio results indicate that the magnetic structure of the alloy is critical in determining its energetics. The ability to model these properties with classic potentials is still to be proven. In this work a detailed comparison between ab initio and classic values of a variety of point defects configurations is performed, testing in this way the extent to which classic potentials can be reliably used for radiation damage studies, and evaluating the dependence of point defect formation energies on Cr concentration.

Thesis. 1978. Ph.D.--Massachusetts Institute of Technology. Dept. of Nuclear Engineering. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / by David C. Aldrich. / Ph.D.

Under simulated accident conditions, tees in the primary coolant loop of a Pressurized Water Reactor (PWR) can deviate from their original design purpose and become separators that effectively remove core heat sink capacity. This method of primary coolant removal is a phenomelogical subset of phase separation known as liquid entrainment, whereby liquid is forced from its original path by the inertia of the gas. A comprehensive literature review revealed common deficiencies in previous studies. The Westinghouse AP600 advanced reactor design was chosen to assess the validity of entrainment models. Following a systematic scaling analysis of the prototypic design a model separate effects test was proposed and constructed at Oregon State University. Just under 100 tests were run to fill the deficiencies found in the literature review. New data from the Air-water Test Loop for Advanced Thermal-hydraulic Studies (ATLATS) could not be predicted by published correlations. A new theoretical model for predicting liquid entrainment onset and steady state entrainment was developed. Comparison with all available data shows a marked improvement for predicting the mass flow rate out the vertical branch. / Graduation date: 2003

A transient code (TFETC) for calculating the temperature distribution throughout the radial and axial positions of a thermionic fuel element (TFE) has been successfully developed. It accommodates the variations of temperatures, thermal power, electrical power, voltage, and current density throughout the TFE as a function of time as well as the variations of heat fluxes arising from radiation, conduction, electron cooling, and collector heating. The thermionic fuel element transient code (TFETC) is designed to calculate all the above variables for three different cases namely: 1) Start-up; 2) Loss of flow accident; and 3) Shut down. The results show that this design is suitable for space applications and does not show any deficiency in the performance. It enhances the safety factor in the case of a loss of flow accident (LOFA). In LOFA, it has been found that if the mass flow rate decreases exponentially by a -0.033t, where t is a reactor transient time in seconds, the fuel temperature does not exceed the melting point right after the complete pump failures but rather allows some time, about 34 seconds, before taking an action. If the reactor is not shut down within 34 seconds, the fuel temperature may keep increasing until the melting point of the fuel is attained. On the other hand, the coolant temperature attains its boiling point, 1057 ��K, in the case of a complete pump failure and may exceed it unless a proper action to trip the reactor is taken. For 1/2, 1/3, and 1/4 pump failures, the coolant temperatures are below the boiling point of the coolant. / Graduation date: 1994

The efficiency of nuclear reactors can be improved by increasing the operating pressure of current nuclear reactors. Current CANDU-type nuclear reactors use heavy water as coolant at an outlet pressure of up to 11.5 MPa. Conceptual SuperCritical Water Reactors (SCWRs) will operate at a higher coolant outlet pressure of 25 MPa. Supercritical water technology has been used in advanced coal plants and its application proves promising to be employed in nuclear reactors. To better understand how supercritical water technology can be applied in nuclear power plants, supercritical water loops are used to study the heat transfer phenomena as it applies to CANDU-type reactors. A conceptual design of a loop known as the Supercritical Phenomena Experimental Apparatus (SPETA) has been done. This loop has been designed to fit in a 9 m by 2 m by 2.8 m enclosure that will be installed at the University of Ontario Institute of Technology Energy Research Laboratory. The loop include components to safely start up and shut down various test sections, produce a heat source to the test section, and to remove reject heat. It is expected that loop will be able to investigate the behaviour of supercritical water in various geometries including bare tubes, annulus tubes, and multi-element-type bundles. The experimental geometries are designed to match the fluid properties of Canadian SCWR fuel channel designs so that they are representative of a practical application of supercritical water technology in nuclear plants. This loop will investigate various test section orientations which are the horizontal, vertical, and inclined to investigate buoyancy effects. Frictional pressure drop effects and satisfactory methods of estimating hydraulic resistances in supercritical fluid shall also be estimated with the loop. Operating limits for SPETA have been established to be able to capture the important heat transfer phenomena at supercritical conditions. Heat balance and flow calculations have been done to appropriately size components in the loop. Sensitivity analysis has been done to find the optimum design for the loop. / UOIT

The flooding phenomenon can be defined as the maximum attainable flow condition beyond which the well defined countercurrent flow pattern can no longer exist. Thus the countercurrent flow limit (CCFL) or the flooding limit may be thought of as the flow condition at which the strong interaction between the two phases occurs. Considerable effort has been devoted to understanding and analyzing the flooding transition in many fields. For example; the flooding phenomenon is one of the important phenomena encountered in the safety analysis of light water reactors (pressurized water reactors and boiling water reactors). Accurate predictions of flooding behavior are particularly important in the assessment of emergency core cooling system (ECCS) performance. Currently, the postulated loss-of-coolant accident (LOCA) is considered the design basis accident. A physical understanding of the flooding phenomenon will help assess core refill during the course of a LOCA. Understanding the physical mechanisms of the flooding phenomenon might help establish more reliable equations and correlations which accurately describe the thermal hydraulic behavior of the system. The models can provide best-estimate capability to the design codes used in the evaluation of ECCS performance. The primary concern of this study was to: 1. Understand the physical mechanisms involved in the flooding phenomenon in order to derive a suitable analytical model. 2. Show that the combination of: a. Linear Instability Theory b. Kinematic Wave Theory c. Catastrophe Theory can provide a general model for flooding phenomenon. The theoretical model derived using the aforementioned combination of theories indicates good agreement between the experimental and the predicted values. Comparisons have been made using a large volume of air-water flooding data. / Graduation date: 1991

Steam interfacial condensation in a core makeup tank was simulated using the code RELAP5/MOD3 version 8.0 to predict the violent pressure oscillation phenomena in a core makeup tank. Six base cases were carried out to study the effects of back pressure and of vacuum conditions produced in the core makeup tank by rapid steam condensation. The effect of varying the liquid conduction thermal layer thickness was studied. In addition, the code's ability to predict condensation heat transfer was evaluated. Violent pressure oscillations were found in the early period of a transient. The violent pressure oscillations had no effect on the total amount of injection water from core makeup tank. The conduction thermal layer thickness was found to only effect the liquid temperature history. The current version of RELAP5/MOD3 was found to be incapable of dealing with the condensation heat transfer problem in which the volume liquid temperature is lower than the temperature of the heat structure which is connected to that hydraulic volume. / Graduation date: 1994

Thesis (Ph. D.)--University of Hawaii at Manoa, 2005. / Includes bibliographical references (leaves 129-135).

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