The structure of turbulence in the subchannels of a large-scale 60 degree section of a CANDU 37-rod bundle was studied at Reynolds numbers equal to 50,000, 100,000 and 130,000. Measurements were conducted at roughly 33.81 rod diameters from the inlet of the rod bundle using single-point, two-component hot-wire anemometry. Analysis of the axial velocity signal indicated a weak effect of Reynolds number on the axial velocity distribution and a bulging of axial velocity contours toward the narrow gaps. The normalised normal Reynolds stresses and the normalised turbulent kinetic energy were found to decrease as the Reynolds number increased. The radial Reynolds shear stress varied linearly with radial distance from the rod, crossing zero at the location of local maximum of the axial velocity. This stress was symmetric about the central rod whereas the azimuthal Reynolds shear stress was anti-symmetric. The Reynolds number effect was weak but measurable on the integral length scales of the axial and radial velocity fluctuations but negligible on the integral length scale of the azimuthal velocity fluctuations, especially in the gap regions. The Taylor and Kolmogorov microscales increased from the wall toward the centre of the subchannel and decreased as the Reynolds number increased. The wall shear stress stress distribution around the central rod indicated no effect of Reynolds number, when normalized by the corresponding average. The wall shear stress reached local minima at rod-wall and rod-rod gaps and local maxima in the open flow regions. Vortex streets were generated within the subchannels very close to the inlet of the rod bundle. The convection speed and frequency of the vortex street were found to increase proportionately to Reynolds number, whereas the vortex spacing was not affected by the Reynolds number.

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)

<p> In this report, the theory for the Improved Quasistatic (IQS) method of solving the three-dimensional, two-neutron-energy group, time-dependent neutron diffusion equations is developed, and approximations appropriate to the CANDU-PHW reactor system are introduced. The theory is extended to a numerical formulation of the problem. The TM-2 computer program (written in FORTRAN 5), which employs the IQS method to numerically solve a two-dimensional form of the diffusion equations (with a correction to account for axial leakage), is outlined. Input and output descriptions for the TM-2 code are provided.</p> / Thesis / Master of Engineering (MEngr)

This is Part A. / <p> Effects in 19 element CANDU fuel bundles containing ThO2 and UO2, located in thermonuclear reactor blankets, have been examined for a variety of blanket designs. The buildup in time of nuclides derived from Th232 by neutron capture was studied. The essential CTR (Controlled Thermonuclear Reactor) blanket features that were examined were tritium breeding in the blanket (the fusion reaction of interest was the DT reaction resulting in the production of a 14.1 MeV neutron and a 3.5 MeV alpha particle), neutron multiplication in the blanket and the U233 production. Means of optimizing these features were also examined. Some conclusions concerning the use and influence of the CANDU bundle are made.</p> <p> It is of interest to study the details of the buildup of U233 in the bundles with the view of perhaps transferring bundles directly from the CTR blanket into a core of a fission power reactor. As such, it would be necessary to convert approximately 3% of the initial thorium present to U233 before transferral. This report examines time steps over which this could be achieved and looks at the performance of the blanket as a whole when such changes occur. The production of unwanted radioactive isotopes is looked at with suggestions of how to minimize this production without harming the rest of the blanket's basic functions.</p> <p> Procedures outlined in the preliminary report "Bench-Mark Neutronic Calculations for Fusion Reactor Designs" by S.A. Kushneriuk and P.Y. Wong form a basis for all of the calculations made in this report. Based on the findings of that report, it is expected that values presented here do reflect, to a fair degree of accuracy, conditions encountered in the CTR blankets studied.</p> / Thesis / Master of Engineering (MEngr)

Under normal operating conditions a CANDU reactor pressure tube (PT) is insulated from its outer calandria tube (CT) by a CO2 gas annulus. If the primary loop coolant flow is compromised the PT can overheat and, if still pressurized, balloon into contact with the CT. At this point the moderator acts as an emergency heat sink. If the heat transferred from the CT to the moderator exceeds the critical heat flux (CHF) the CT can overheat, begin to strain due to the contact pressure, and eventually fail. A mechanistic model is presented that describes ballooning contact of the PT and CT, the resulting thermal contact conductance, heat flux to the moderator, and, if CHF is exceeded, the development of film boiling and potential CT strain. The goal is to create a software package that predicts fuel channel failure during a pressure tube overheat event. / Thesis / Master of Applied Science (MASc) / Computer software was developed to predict CANDU fuel channel failure in the event of a total station blackout. The model created successfully predicted the available experimental data.

This work can be viewed in three separate sections, each of which build off of the prior. The first part of this study examined the flow in a 1/16th scale calandria test section based on a typical CANDU moderator layout. The experiments utilized forced flow supplied to the vessel and electrical heated rods to mimic the heat flow from calandria tubes. The size of the vessel, flow rates, and power levels were used to scale the experiments such that the provided representative temperature fields. The temperature field inside the vessel was measured and shown to compare well with CFD predictions over a wide range of inlet conditions and power levels. Additionally, this work addressed the scaling distortions in the experiment which occurred due to physical limitations when performing experiments at 1/16 scale (e.g., a smaller number of heater rods with a larger diameter were used in the experiment because at 1/16-scale direct fabrication of 390 fuel channel simulators is not feasible). The work proposed the H factor addition to the Ar. This additional scaling criteria was shown to better maintain the flow regimes expected CANDU moderators by taking into account distortions introduced by surface heating instead of volumetric heating in addition to the reduction in total number of tubes. While this work involved forced convective flows at the inlet of the vessel, in some regions of the calandria buoyancy induced forces were sufficiently high such that these phenomena altered the direction and magnitude of the flows as compared to purely forced convective behavior. Hence further work, discussed below, was initiated to better understand and measure these local phenomena where buoyancy forces are of similar magnitude as those of forced convection. Such local conditions we have terms mixed convection regime for the purposes of this thesis. The second part of this work further examined the mixed convection between a subset of the CANDU calandria tubes, namely how does a lower tube effect the mixed convection heat transfer of the upper tube in an inline arrangement. To isolate and measure the phenomena with sufficient detail, a small number of tubes was studied and advanced diagnostics such as Particle Image Velocimetry (PIV) and Laser Induced Fluorescence (LIF) were employed. This study combined fluid velocity, temperature and wall temperature measurements with CFD simulations to develop a mechanistic model and understanding of the effect of natural convection plumes from lower elevations on the natural circulation phenomena on an upper cylinder. Superposition of the natural convection phenomena combined with pseudo forced convection effects from the lower elevation cylinder’s plume was used to model the mixed convection phenomena. This model was shown to perform well, with nearly all data being predicted to with +-20% for experiments performed in this work, and experiments in literature. A major finding from the preceding discussion is the importance of the lower elevation plume velocity on the local phenomena on the upper cylinder. The third section further expanded upon the prior two by replacing the lower cylinder with a diffuser nozzle which could provide a forced convective component with accurately defined velocities. Such measurements allow for accurate definition of the local Ri number and allowed full access for instrumentation to observe the velocity fields. The major contribution of this work was a flow regime map that defined the phenomena around a heated cylinder under mixed convection conditions. Additionally, the establishment of a database of fluid temperature and velocity measurements for a wide range of Ri was also developed and used to further validate CFD predictions. / Thesis / Doctor of Philosophy (PhD)

<p>The CANDU-900 reactor design is an improvement on the current CANDU-6 reactor in the areas of economics, safety of operation and fuel cycle flexibility. As power grids start to rely more heavily on nuclear, it will be imperative for future nuclear generating station designs to be able to adjust their output to suit the fluctuating demands of the grid. Additionally, the need to reduce global nuclear waste has motivated research into mixed oxide fuel with the goal of maximizing spent fuel repository capacity and reducing decay heat via transmutation of transuranic actinides. The objective of this thesis is to provide insight into the load following capabilities of the CANDU-900 reactor design for a transuranic mixed oxide (TRUMOX) fueled core.</p> <p>The three-dimensional fuel management code, RFSP-IST, was used to simulate a reactor operating history for week long load following operations in a generic CANDU-900 reactor. Daily refuelling operations as well as reactivity device movements supplementary to RFSP were performed using the RECORD RRS emulator program. Core snapshots were taken at periodic intervals using the SIMULATE module to observe and track various reactor parameters. Average liquid zone controller fills as well as core reactivity and channel power values were used to determine the controllability of the reactor for various load following depths.</p> <p>The results of the load following simulations show that TRUMOX fuel has superior load following capabilities to that of conventional NU fuel for practical operational scenarios in a CANDU-900 reactor. Load following operations could be performed for TRUMOX fuel down to 85% full power in a safe and controllable manner using only the liquid zone controllers to account for the xenon transient reactivity as compared to NU which could only be done down to 90% full power. For load following simulations that both fuel types were capable of performing in a controllable manner, the TRUMOX fuelled core maintained on average a larger safety margin between the average liquid zone controller fills and the established safety limits.</p> / Master of Applied Science (MASc)

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)

Aujourd’hui, l’énergie nucléaire représente une partie non-négligeable du marché énergétique mondial, très probablement vouée à croître dans les prochaines décennies. Les réacteurs du futur devront notamment répondre à des critères supplémentaires économiques mais surtout de sûreté, de non-prolifération, de gestion optimisée du combustible et d’une gestion responsable des déchets nucléaires. Dans le cadre de cette thèse, des études concernant la non-prolifération des armes nucléaires sont abordées, dans le cadre de la recherche et développement d’un nouvel outil potentiel de surveillance des réacteurs nucléaires ; la détection des antineutrinos des réacteurs. En effet, les propriétés de ces particules pourraient intéresser l’Agence Internationale de l’Energie Atomique (AIEA) en charge de l’application du Traité de non-prolifération des armes nucléaires. L’AIEA encourage ainsi ses états membres à mener une étude de faisabilité. Une première étude de non-prolifération est réalisée avec la simulation d’un scénario proliférant utilisant un réacteur de type CANDU et de l’émission en antineutrinos associée. Nous en déduisons une prédiction de la sensibilité d’un détecteur d’antineutrinos de taille modeste à la diversion d’une quantité significative de plutonium. Une seconde étude est réalisée dans le cadre du projet Nucifer, détecteur d’antineutrinos placé auprès du réacteur de recherche OSIRIS. Nucifer est un détecteur d’antineutrinos dédié à la non-prolifération à l’efficacité optimisée conçu pour être un démonstrateur pour l’AIEA. La simulation du réacteur OSIRIS est développée ici pour le calcul de l’émission d’antineutrinos qui sera comparée aux données mesurées par le détecteur ainsi que pour caractériser le bruit de fond important émis par le réacteur détecté dans Nucifer. De façon générale, les antineutrinos des réacteurs sont émis lors des décroissances radioactives des produits de fission. Ces décroissances radioactives sont également à l’origine de la puissance résiduelle émise après l’arrêt d’un réacteur nucléaire, dont l’estimation est un enjeu de sûreté. Nous présenterons dans cette thèse un travail expérimental dont le but est de mesurer les propriétés de décroissance bêta de produits de fission importants pour la non-prolifération et la puissance résiduelle des réacteurs. Des premières mesures utilisant la technique de Spectroscopie par Absorption Totale (TAGS) ont été réalisées auprès du dispositif de l’Université de Jyväskylä. Nous présenterons la technique employée, le dispositif expérimental ainsi qu’une partie de l’analyse de cette expérience. / Today, nuclear energy represents a non-negligible part of the global energy market, most likely a rolling wheel to grow in the coming decades. Reactors of the future must face the criteria including additional economic but also safety, non-proliferation, optimized fuel management and responsible management of nuclear waste. In the framework of this thesis, studies on non-proliferation of nuclear weapons are discussed in the context of research and development of a new potential tool for monitoring nuclear reactors, the detection of reactor antineutrinos, because the properties of these particles may be of interest for the International Agency of Atomic Energy (IAEA), in charge of the verification of the compliance by States with their safeguards obligations as well as on matters relating to international peace and security. The IAEA encouraged its member states to carry on a feasibility study. A first study of non-proliferation is performed with a simulation, using a proliferating scenario with a CANDU reactor and the associated antineutrinos emission. We derive a prediction of the sensitivity of an antineutrino detector of modest size for the purpose of the diversion of a significant amount of plutonium. A second study was realized as part of the Nucifer project, an antineutrino detector placed nearby the OSIRIS research reactor. The Nucifer antineutrino detector is dedicated to non-proliferation with an optimized efficiency, designed to be a demonstrator for the IAEA. The simulation of the OSIRIS reactor is developed here for calculating the emission of antineutrinos which will be compared with the data measured by the detector and also for characterizing the level of background noises emitted by the reactor detected in Nucifer. In general, the reactor antineutrinos are emitted during radioactive decay of fission products. These radioactive decays are also the cause of the decay heat emitted after the shutdown of a nuclear reactor of which the estimation is an issue of nuclear safety. In this thesis, we present an experimental work which aims to measure the properties of beta decay of fission products important to the non-proliferation and reactor decay heat. First steps using the technique of Total Absorption Gamma-ray Spectroscopy (TAGS) were carried on at the radioactive beam facility of the University of Jyvaskyla. We will present the technique used, the experimental setup and part of the analysis of this experiment.

Inconel alloys are commonly used as structural materials in nuclear reactors. One of these alloys, the Inconel X-750, is a γ’ Ni3(Al, Ti) strengthened superalloy extensively used in the cores of reactors, such as spacers in CANada Deuterium Uranium (CANDU) fuel channels. Prior to their application in commercial reactors, accelerated irradiation tests had been conducted in liquid metal fast reactors. Results did not indicate any problem stemming from significant fast neutron irradiation. However, recently it has been found that the ex-service CANDU Inconel X-750 spacers became severely brittle after a lengthy exposure to reactor environment. The underlying mechanism remains unclear and thus forms the focus of this current investigation, predominantly through transmission electron microscopy (TEM). This dissertation unfolds with the literature review in Chapter 2, followed by presentation of novel techniques in Chapter 3 on the preparation of TEM samples from small reactor components, namely the spacers. Chapter 4 presents TEM characterizations of ex-service spacers removed from the reactors. To simulate neutron irradiation over wide temperature range in an effort to understand the damage mechanisms, heavy ion irradiations were conducted and reported in Chapter 5 and 6. Irradiations are found to significantly alter the stability of the primary strengthening phase γ’, a systematic experimental study of which is presented in Chapter 7. To fully understand the effects of transmutation produced helium on irradiation induced cavity and dislocation microstructures, TEM in-situ heavy ion irradiations with hot/cold pre-injected helium were conducted and reported in Chapter 8 and 9. Helium was found to play an important role in the irradiation-induced instability of γ’ in nickel-based superalloys, the discussion of which is presented in Chapter 10. As one of the most important defect structures induced from irradiation, the stacking-fault-tetrahedra, were dynamically observed and are described in a journal manuscript in Appendix A. In addition to broadening current understanding of material degradation mechanism for in-service CANDU spacer, this study also provides comprehensive information on irradiation damage in nickel based superalloys, irradiation induced lattice defects and phase instability in face centered cubic alloys, as well as helium’s effects on cavity formation, dislocation evolution, and phase transformation. / Thesis (Ph.D, Mechanical and Materials Engineering) -- Queen's University, 2013-09-06 15:21:02.334