The present work focuses on the study of two-phase flow behavior called “Intermittent Buoyancy Induced Flow” (IBIF) resulting from the loss of coolant circulation in a CANDU nuclear reactor core. The main objectives are to study steam bubble formation and migration through the pressure tube and into the feeder tubes and headers, and to study the effect of pressure tube sagging on the two-phase flow behavior during IBIF. Experiments are conducted using air and water flow at atmospheric pressure to qualitatively examine the IBIF phenomena. The test showed oscillating periodic behavior in the void fraction as the air vents. In addition to this, a mathematical model based on a simplified momentum balance for the liquid and gas phases was formulated. The model was further solved and compared to the experimental data. The model predictions showed a reasonable agreement within the investigated range of void fractions.

The present work focuses on the study of two-phase flow behavior called “Intermittent Buoyancy Induced Flow” (IBIF) resulting from the loss of coolant circulation in a CANDU nuclear reactor core. The main objectives are to study steam bubble formation and migration through the pressure tube and into the feeder tubes and headers, and to study the effect of pressure tube sagging on the two-phase flow behavior during IBIF. Experiments are conducted using air and water flow at atmospheric pressure to qualitatively examine the IBIF phenomena. The test showed oscillating periodic behavior in the void fraction as the air vents. In addition to this, a mathematical model based on a simplified momentum balance for the liquid and gas phases was formulated. The model was further solved and compared to the experimental data. The model predictions showed a reasonable agreement within the investigated range of void fractions.

Part A of two parts. Part B titled: "Xenon Transient Studies For a CANDU Reactor". / <p> In CANDU reactors, mechanical devices called shutoff rods are used to shutdown the reactor if required. These rods are made of high thermal neutron absorbing material such as cadmium. The number and the locations of the shutoff rods are optimized for a given reactor configuration. Optimization here means minimizing the number of rods and maximizing their reactivity depth or effectiveness. </p> <p> Optimization may be studied in various ways but the method selected is both simple and basic. It is apparent that if the interaction effects between the individual shutoff rods are reduced, their worths will increase. The optimum distance between two rods was determined to be 130 cm. Also, the best location of a third rod with respect to two already placed at an optimum separation was studied. Finally, these results were used in order to determine the optimum distance between banks of shutoff rods. These banks of rods were arranged in such a way as to achieve maximum flux flattening with all the rods inserted in the core. A 22 shutoff rod configuration for an adjuster flattened CANDU reactor gave a total change of 5.6% in keff. </p> / Thesis / Master of Engineering (MEngr)

Part B of two parts. Part A found at: http://hdl.handle.net/11375/18745 / <p> This report studies the xenon transient behaviour in a CANDU reactor as a function of time after shutdown, start-up and power setbacks. In addition, load cycling transients were obtained for typical daily load requirements. </p> / Thesis / Master of Engineering (MEngr)

This file is officially titled as a project. This is Part B in connection with another project by the same author, titled "Part A: Adjuster Rod Design in a CANDU Reactor and Flux Distributions Due to an Arbitrary Source of Neutrons."http://hdl.handle.net/11375/15387 / In this report, the effectiveness of a proposed Shutoff Rod System a CANDU Reactor was investigated. A full core simulation was done, to study the neutronic power transient following a change in coolant conditions. / Thesis / Master of Engineering (ME)

<p> In a postulated critical break LOCA in a CANDU reactor it is possible that heatup of a pressure tube (PT) causes ballooning contact with the calandria tube (CT). Stored heat in the PT is transferred out, yielding a high PT-CT heat flux, which can cause dry out of the CT and establishment of pool film boiling on the outer surface of the tube. The safety concern associated with this condition is that if the temperature of the CT experiencing film boiling gets sufficiently high then failure of the fuel channel may occur. However, quench heat transfer can limit the extent and duration of film boiling as has been experimentally observed. Current estimates of quench temperatures during pool film boiling are based primarily on experimental correlations. In this dissertation a novel mechanistic model of pool film boiling on the outside of a horizontal tube with diameter relevant to CT (approximately 130 mm) has been developed. The model is based in part upon characterizing the vapor film thickness for steady state film boiling under buoyancy driven natural convection flows around a tube located horizontally in a large liquid pool. Variations in steady state vapor film thickness as a function of the incident heat flux, the temperature of the CT outer wall, and the subcooling of the bulk liquid are analyzed. The calculated effective film boiling heat transfer coefficient is compared to available experimental data. Finally a transient equation is developed which quantifies the instability of the vapor film and a possible occurrence of rapid quench when a step change in governing parameters occurs, such as liquid subcooling. This mechanistic model can be employed in safety analysis to demarcate the conditions under which fuel channel failure will not occur in a postulated critical break LOCA.</p> / Thesis / Master of Applied Science (MASc)

Estimates of fuel management data for the equilibrium operation of a specific CANDU reactor have been obtained by simulating a period of the reactor's history using the Fuel Management Design Program (FMDP). The collapsing procedure in FMDP has been tested and improved. This procedure prepares a coarse mesh model of the reactor core from a detailed fine mesh calculation. The program calculates a set of coarse mesh parameters which, when used in the flux calculation, will regenerate exactly the same eigenvalue and flux distribution as the fine mesh model. These parameters can then be used with the coarse mesh, to calculate flux distributions for a series of perturbations from the reference calculation used in collapsing. Several coarse mesh models were generated and studied. It was found that coarse mesh calculations with collapsed parameters result in large savings in computing costs compared to the same calculations with fine mesh, but with very little loss in accuracy. / Thesis / Master of Engineering (ME)