<p>The reprocessing of spent fuel is a key component in reducing the end waste from nuclear power plant operations and creating a sustainable closed fuel cycle. Central to this effort is the extraction and reprocessing of actinide materials to be recycled into fast or thermal reactors. Reprocessed actinides can contribute additional energy and may be partially transmuted in current thermal systems using mixed oxide fuels before being sent to fast reactors. The use of current thermal reactors as an intermediary step significantly reduces the fast reactor infrastructure needed to handle the spent fuel inventory in the long term, and also provides a source of additional energy from existing mined resources in the short term. An optimization of the fast and thermal systems in a closed fuel cycle reduces the end cycle waste to primarily fission products which have little residual value and manageable disposal and monitoring demands. The dissertation explores the design and analysis of an actinide transmutation solution utilizing a current thermal reactor design. The TRUMOX-30 CANDU-900 system defined herein uses a mixed oxide fuel containing 3.1% transuranic actinides extracted from 30 year cooled spent fuel from a prototypical Pressurized Water Reactor (PWR) and mixed with natural uranium. A significant constraint imposed on the design is that the actinide burning is to occur in an existing CANDU design without major changes or infrastructure replacement. Hence the standard CANDU design and analysis methodology was employed to produce and evaluate the system. The phased approach includes extensive neutron transport modeling of the lattice and control device super-cell configurations, which feed forward in to a detailed full core diffusion model of the TRUMOX-30 CANDU-900 design. Suitable fuel burnup and significant actinide conversion was achieved while remaining within the prescribed operational envelope of the CANDU reactor. The design was evaluated against existing operational constraints and limits, performing well and achieving the goal of actinide transmutation with no changes to the reactor design. This effort demonstrated the adaptation of a current CANDU-900 reactor as a platform for intermediary actinide transmutation which may form part of a sustainable and efficient fuel cycle.</p> / Doctor of Philosophy (PhD)

<p>When simulating reactor physics experiments, uncertainties in nuclear data result in a bias between simulated and experimental values. For new reactor designs or for power reactor designs the bias can be estimated using a set of experiments. How- ever, the experiments used to estimate the bias must be applicable to the power reactor design of interest. Similarity studies can be performed to ensure this is the case. Here, potential experiments in the ZED-2 heavy water critical facility at Chalk River Laboratories were developed that would be applicable to the multiplication factor bias calculation of three thoria plutonia fuelled power reactor designs. The power reactor designs that were analyzed were the CANDU 6 with 37-element fuel bundles and 43-element fuel bundles, and a Canadian SCWR design with 78- element fuel assemblies. The power reactors were simulated using the code package SCALE 6.1 under burnup conditions that were determined using the lattice code DRAGON 3.06H and the diffusion code DONJON 3.02A. The intermediate results from DRAGON and DONJON were used to compare the benefits of various reactor designs. Various critical core configurations were then simulated in the ZED-2 re- actor using the SCALE 6.1 package. The similarities between the potential ZED-2 reactor experiments and the power reactors were analyzed. These results were used to design a set of experiments having sufficiently high completeness that they can be used as part of a bias calculation using the generalized linear least squares method. To do so a methodology was developed to guide the experiment set design process in which the fuel type, lattice arrangements, and coolant type are modified and the effects on the sensitivity coverage analyzed. A set of six experiments was designed for which all of the power reactor designs had a completeness of 0.7 or higher.</p> / Master of Applied Science (MASc)

Study of quench cooling is very important in nuclear reactor safety for limiting the extent of core damage during the early stages of severe accidents after Loss of Coolant Accidents (LOCA). Quench of a hot dry surface involves the rapid decrease in surface temperature resulting from bringing the hot surface into sudden contact with a coolant at a lower temperature. The quench temperature is the onset of the rapid decrease in the surface temperature and corresponds to the onset of destabilization of a vapor film that exists between the hot surface and the coolant. Re-wetting the surface is the establishment of direct contact between the surface and the liquid at the so-called re-wetting temperature. Re-wetting is characterized by the formation of a wet patch on the surface which then spreads to cover the entire surface. Situations involving quench and re-wetting heat transfer are encountered in a number of postulated accidents in Canada Deuterium Uranium (CANDU) reactors, such as re-wetting of a hot dry calandria tube in a critical break LOCA. This accident results in high heat transfer from the calandria tube to the surrounding moderator liquid which can cause the calandria tube surface to experience dryout and a subsequent escalation in the surface temperature. If the calandria tube temperature is not reduced by initiation of quench heat transfer, then this may lead to subsequent fuel channel failure. In literature very limited knowledge is available on quench and re-wetting of hot curved surfaces like the calandria tubes. In this study, a Water Quench Facility (WQF) has been constructed and a series of experiments were conducted to investigate the quench and re-wetting of hot horizontal tubes by a vertical rectangular water multi-jet system. The tubes were heated to a temperature between 380-800°C in a controlled temperature furnace then cooled to the jet temperature. The temperature variation with time in the circumferential and the axial directions of the tubes has been measured. The twophase flow behavior and the propagation of the re-wetting front around and along the tubes were simultaneously observed by using a high-speed camera. The effects of several parameters on the cooling process have been investigated. These parameters include: initial surface temperature, water subcooling (in the range 15- 800C), jet velocity (in the range 0.15-1.60 m/s), tube solid material (brass, steel and Alumina), surface curvature, tube wall thickness, jet orientation and number of jets. The variables studied include the re-wetting delay time (time to quench after initiating the cooling process), there-wetting front propagation velocity, the quench and re-wetting temperatures, the quench cooling rates and the boiling region size. The quench and the re-wetting temperatures as well as the re-wetting delay time were found to be a strong function of water subcooling. The quench and re-wetting temperatures increase with increasing water subcooling. The rewetting delay time decreases with increasing the water subcooling, decreasing initial surface temperature, increasing liquid velocity and decreasing the surface curvature. There-wetting front velocity is mainly dependent on the initial surface temperature and water subcooling. The re-wetting velocity increases by decreasing the initial surface temperature and by increasing the water subcooling. Decreasing the surface curvature was found to also increase the re-wetting front velocity. Correlations of the phenomena studied have been developed and provided good prediction of the experimental data collected in this study and data available from literature. The. results of this study provide novel knowledge and an experimental database for mechanistic modeling of quench heat transfer on calandria tube surfaces that experience dryout and film boiling. / Thesis / Doctor of Philosophy (PhD)

Fuel comprised of thorium has been explored since the early development of nuclear energy in the 1960s. In the last decade, there has been a renewed interest in thorium fuel and it has now become a primary focus in studies and proposed in next-generation nuclear reactors. This has been prompted by a limited supply of uranium in the foreseeable future and an abundance of thorium resources. Additionally, when compared to natural uranium (NU), thoria (ThO$_2$) produces substantially less long-lived radioactive waste and the fissile content can be reprocessed for additional fuel cycles. The CANDU 6 reactor has a unique ability to harvest thorium fuels because of its superior neutron economy. Thorium requires a driving isotope to sustain neutron fission until the long absorption chain produces viable amounts of U-233. Previous studies have investigated many different practical fissile isotopes and core modeling techniques that would make thorium feasible in a CANDU 6 reactor. This thesis focuses on a safety evaluation of thorium fuel compared to NU fuel in a lattice cell and full core configuration. \\ DRAGON 3.06 and SERPENT 2 are used to examine the infinite lattice cell containing NU and homogeneous thorium fuel enriched with 2.0\% U-235, emphasizing the relationship between multiple nuclear libraries. This configuration is used to determine the enrichment concentration, temperature coefficient, coolant void reactivity, and the power relationship. Thorium fuel exhibits a higher negative temperature coefficient, a lower coolant void reactivity, and a greater reactivity change when simulated at different powers. If the lattice cell is simulated at 75\% nominal power there is an 11 mk adjustment for thorium fuel, whereas the adjustment is only three mk for NU fuel. This is related to the extensive cross section of Th-232 and the long fertile absorption chain results in a sizeable inventory of the intermediate isotope Pa-233. The fissile content of the fuel bundle after exiting the reactor will continue to accumulate U-233 and should be monitored and properly stored. \\ A full core evaluation in a CANDU 6 reactor is performed in DONJON 4. Thorium fuel has an inferior reactivity worth for the control mechanism than does NU fuel in an operating CANDU 6 reactor. The reactivity worth of leakage and absorption in the reactor is estimated to be slightly lower for the thorium fuel. \\ This thesis presents a new computational model for analyzing full core power transients built upon previous results. The approximation model utilizes many assumptions to develop an expeditious code for analyzing the infinite square lattice retaining the isotopic densities. This model has demonstrated the ability to accurately emulate the reactivity of a lattice cell at different powers and power transients formed in DRAGON. The model is coupled with a point kinetic code to perform power transients in a CANDU 6 reactor. \\ Load following operations are performed in cycles of 24 hours examined at 80\%, 60\%, and 40\% full power. Power adjustments are performed in increments of 10 minutes, two hour, or four hour periods with a constant reactivity input. The power adjustment time has minimal effect on the reactivity perturbations and only influences the rate of reactivity. Thorium fuel has enhanced load following capabilities compared to conventional NU fuel.\\ The long-lasting effects of Pa-233 introduces safety concern when reducing power or reactor shutdown scenarios. Reactivity transformation within the first two days of immediate power reduction will yield similar results for both fuels. Excess reactivity in the thorium fuel will continue to accumulate and eventually double the reactivity peak of NU fuel in the following 90 to 120 days. A shutdown simulation is performed in incremental power reduction steps of 20\% for a range of different days. It is found that NU fuel can adequately control the additional reactivity in this simulation. Thorium fuel maintains a disconcerting amount of excess reactivity that will need to be addressed accordingly. The protactinium transient highlights the need to adequately monitor the buildup of Pa-233 for thorium-based fuels in a reactor. / Thesis / Master of Applied Science (MASc)

McMaster University MASTER OF APPLIED SCIENCES (2017) Hamilton, Ontario (Engineering Physics) TITLE: Comparative study of accident-tolerant fuel for a CANDU lattice AUTHOR: Simon Younan, B.Eng. (McMaster University) SUPERVISOR: Dr. David Novog NUMBER OF PAGES: xiii, 120 / Since the Fukushima accident in 2011, there have been an increasing number of studies on the use of accident-tolerant fuel (ATF) in light water reactors to mitigate the consequences of a future severe accident, by better retaining fission products and/or providing operators more time to implement emergency measures. However, few studies exist for CANDU reactors in this regard. The goal of this study is to determine how different types of ATF are expected to behave in a CANDU lattice when compared to the current UO2 fuels. In particular, this study focuses on neutronic parameters calculated using the Serpent 2 code, but also models heat transfer and stylized accident scenarios. The ATF concepts tested include UO2-SiC composites, UN and UN-based composites, U-9Mo, and fully ceramic microencapsulated (FCM) fuel, along with SiC and SS-coated cladding. Four general conclusions can be drawn: 1. Fuel temperature are lower for ATF as compared to traditional fuels. UO2-SiC composite fuel exhibits a moderate temperature reduction compared to UO2, particularly for fresh fuel. Other ATF fuel materials exhibit a substantial decrease in fuel temperature compared to UO2. The lower fuel temperatures are also accompanied by lower melting temperatures for some fuels, hence each design requires specific assessments on safety. 2. As most ATF have a poorer neutron economy compared to standard fuel designs, enrichment is required to use ATF in a CANDU, particularly for UN and FCM fuel compositions. Coolant void reactivity (CVR) is lowest with FCM fuel and highest with U-9Mo fuel. Fuel temperature coefficient (FTC) is most negative for fuel containing UN or U-9Mo. 3. Changing the cladding material from zircaloy to SiC slightly improves neutron economy, while a FeCrAl surface layer impairs neutron economy. The impact of many ATF sheath materials is to greatly reduce or eliminate hydrogen production in some severe accidents. A specific assessment on hydrogen production was not performed in this study. 4. In stylized accident scenarios, all fuels exhibit only a small temperature spike due to the reactivity insertion of the LOCA as the reactor shutdown limits the power excursion. For cases where Emergency Core Cooling functions as designed, fuel and channel failures are precluded for both traditional fuels and ATF. For cases with impairment of ECC, most ATF fuels show lower fuel temperatures than UO2 fuels and adequate heat removal to the pressure-calandria tube fuel channel. The exception would be Mo-based fuels that reach the melting point prior to establishing an adequately high sheath temperature to sustain radiative heat removal to the PT-CT assembly. / Thesis / Master of Applied Science (MASc) / Since the Fukushima accident in 2011, there have been an increasing number of studies on the use of accident-tolerant fuel in nuclear reactors to mitigate the consequences of a future severe accident, reducing the likelihood and severity of a radiation release. Canadian reactors are of the CANDU design, which differs greatly from the reactors most recent studies have focused on. The goal of this study is to determine the feasibility of using accident-tolerant fuel in CANDU reactors, studying different types. In general, the goal of accident-tolerant fuels in CANDU reactors would be to reduce fuel temperatures and improve fission product retention, reducing the likelihood/magnitude of radioactive releases in a severe accident. However, nearly all types of accident-tolerant fuel would also require the uranium to be slightly enriched as opposed to the current fuel which is based on naturally-occurring uranium. This study outlines the results obtained by computer modelling of accident-tolerant fuel in a CANDU reactor, including the enrichment requirements, changes to important reactivity feedbacks, and impacts on accident performance.

<p> Best estimate and analysis of uncertainty methods are used to examine the variability of the H factor, which converts the global flux in a lattice cell to power. The assumption of a constant H factor is tested, by examining the sensitivity of the components of the H factor to perturbations in reactor conditions, such as, moderator temperature, boron content in the moderator, moderator purity, RIH temperature, ROH pressure, and exit burnup. The local flux profile, which is a component of the H factor, is calculated for a typical CANDU reactor lattice cell using WIMS 2.5d. Another component of the H factor, the distribution of fission energy in a lattice cell, is found by exploring the location of each source of energy released from a fission event. To examine the location of the gamma ray energy deposition a two dimensional Monte Carlo code was created and subsequently benchmarked against an analysis done by C.R.Boss. Using the Monte Carlo code, the best estimate of the percentage of gamma ray energy deposited in the heat transport system was found to be 83.7%. The moderator temperature and the exit channel burnup are shown to have the largest influence on the H factor, which was found to vary between 99.6% and 100.4% of the best estimate value.</p> / Thesis / Master of Applied Science (MASc)

<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)

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.

A novel design for an automated uranium pellet stacking system is presented. This system is designed to replace the manual method for stacking uranium pellets for CANDU fuel bundles that is currently used at Cameco Fuel Manufacturing in Port Hope, ON. The system presented is designed as a drop-in solution to the current production line at Cameco. As a result, there are constraints that prevent certain parameters from modification. The three main goals of this system are to reduce worker exposure to radiation to as low as reasonably achievable, improve product quality, and increase the productivity of the production line. The proposed system will remove the workers from a position of having to handle the uranium pellets and physically place them on the stacks. While the natural uranium currently in production is not a major health risk for short-term exposure, the possibility of production of slightly enriched uranium bundles makes this system a real need. This system also removes the random pellet placement that the manual system uses by taking precise measurements using laser triangulation sensors. These measurements are used to determine which sizes of end pellets are required to complete the stack to within the specified tolerances. A final measurement is done to ensure the stack is within tolerance. All of this information is recorded and can be traced back to the stacks during quality inspection, which is a major improvement over the existing system. This single automated system will replace two manual stations, while increasing the total output production, thus eliminating pellet stacking as a bottleneck in the fuel bundle assembly process. Current production rates can be met by this single, automated station in two shifts per day, while the current manual process requires three shifts using two stations. Test results of a proof-of-concept prototype indicate that the proposed design meets or exceeds all of the design requirements.