The Generation IV International Forum has established several goals for the next generation of nuclear energy systems, which are to be substantial improvements over contemporary designs. In Canada Generation IV research efforts have focused on developing the Pressure Tube type SuperCritical Water-cooled Reactor (PT-SCWR), an evolution of CANada Deuterium Uranium (CANDU) technology. An integral part of the PT-SCWR is the High Efficiency Re-entrant Channel (HERC), wherein coolant first travels downward through a centre flow tube and then upward around the fuel. The large density variation of supercritical fluids, combined with the negative Coolant Void Reactivity (CVR), make the concept similar to a Boiling Water Reactor (BWR). The objective of this study was thus to apply the state-of-the-art in BWR analysis to the PT-SCWR. Models were created using the DRAGON (neutron transport), DONJON (neutron diffusion/spatial kinetics), and CATHENA (channel thermalhydraulics) computer codes. A procedure for DONJON-CATHENA coupling was developed to enable simulation of coupled transients. The specifications of the HERC necessitated multiple coolant reactivity feedbacks be included in the model, in turn requiring extensions to the DONJON source code. The model created for this work is thus among the first to incorporate multiple coolant feedbacks in core-level coupled spatial kinetics and thermalhydraulics transient analysis, and is uniquely capable of simulating such transients in the PT-SCWR. This work found that while the total CVR was negative as required, the reactivity effect of coolant void solely around the fuel was positive. As a consequence additional heat delivered from fuel to coolant, which decreases the coolant density, has a positive reactivity effect making BWR-like coupled instabilities impossible. On the other hand, in some postulated transients, such as Loss-OfCoolant Accidents (LOCAs) or Loss-Of-Flow Accidents (LOFAs), this positive reactivity results in temporary power excursions. A fast-acting shutdown system is potentially necessary to limit damage to the fuel in such transients. / Dissertation / Doctor of Philosophy (PhD)

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.

<p>The nuclear safety industry makes extensive use of thermalhydraulics system analysis and computational fluid dynamics codes for validation and predictive purposes. These codes take different approaches to provide the user with reasonable estimates of system and component behaviors. With each displaying its own strengths, it is only logical to pursue coupled systems of these codes to create increasingly accurate, versatile, and more computationally efficient safety analysis tools. This work presents results of the attempted coupling of CD-ADAPCO's STAR-CCM+, a computational fluid dynamics (CFD) code, to Atomic Energy of Canada's CATHENA thermalhydraulics (TH) code. This coupled system is used in the simulation of the conditions within an inlet header of the RD-14M experimental facility under single phase conditions in the initial phase of selected test. This inlet header is removed from a modified CATHENA test B9401 deck and instead modelled in STAR-CCM+. Custom applications were written to allow information exchange at the newly created boundaries to provide an attempt at a coupled system. Results are provided through multiple stages of development of the coupled system, from the unmodified B9401 test case of CATHENA into a coupled system with header behavior predicted by STAR-CCM+. Though successful information transfer between codes was established at each desired time step and interval, the current technique was found to be insufficient for establishing acceptable steady-state conditions for the commencement of more complex (transient and two-phase) conditions.</p> / Master of Applied Science (MASc)

Les réacteurs de 4ème génération ont pour objectif l’avènement d’un nucléaire durable susceptible de soutenir la transition énergétique. Anticipant un possible retard, dû à des difficultés techniques et économiques, des solutions innovantes inspirées des technologies actuelles (REP) sont à l’étude. Ces réacteurs à haute conversion nécessitent le développement d’outils académiques simples et robustes adaptés aux phases de la conception et capables :- D’évaluer les performances du combustible (burnup). Cet aspect est géré par l’environnement C++ SMURE (Serpent/MCNP Utility for Reactor Evolution), ici adapté et complété pour modéliser l’évolution du combustible à l’échelle de l’assemblage comme à l’échelle du cœur.- D’évaluer les performances de sûreté, ce qui nécessite le couplage entre la neutronique, ici approximée par la théorie de la diffusion et résolue par la NDM (Nodal Drift Method), et la thermohydraulique dont le traitement est simplifié dans le code BATH (Basic Approach to ThermalHydraulics). Le couplage NDM/BATH a fait l’objet d’une validation sur un benchmark d’éjection de grappe.Nos outils et méthodes de conception sont appliqués au dimensionnement de SMR sous-modérés à eau légère fonctionnant soit au Th/U soit au Th/Pu. Outre les critères usuels de conception (i.e. facteur de forme), il s’est avéré nécessaire, pour la crédibilité du concept, de spécifier la gestion de la réactivité, ce qui a mené au développement d’une méthodologie d’optimisation des poisons consommables. L’analyse de sûreté a permis de poser de nouveaux critères de conception, notamment sur le niveau maximal de sous-modération permettant d’éviter la crise d’ébullition nucléée. Cela a également mis en lumière les implications sur la sûreté de certains choix de conception comme le fonctionnement avec un inventaire réduit en bore soluble. / The Generation IV of nuclear reactors aims at making the nuclear energy a sustainable power source, able to contribute efficiently to the energetic transition. To anticipate the delay of this Gen. IV, innovative retro-fitted nuclear reactors with high level of conversion are studied. The conception of such reactors needs the development of a flexible and robust academical tool box in order to:- Evaluate fuel performance. This is done by means of SMURE (Serpent/MCNP Utility for Reactor Evolution), the dedicated CNRS C++ framework, which is adapted to perform burnup calculation both at assembly scale and at core scale.- Evaluate safety performance. This implies coupled transient simulation between neutronics and thermohydraulics. Neutronics is handled by the Nodal Drift Method (NDM) which solves the diffusion equations while thermohydraulics is simplified and computed by the code Basic Approach to ThermalHydraulics (BATH). This coupling between NDM/BATH has been validated on a Rod Ejection Accident (REA) benchmark.Ours tools and methods are applied on the design of sub-moderated water-cooled SMR cores using either Th/U or Th/Pu fuel. In addition to basic conception criteria such as the form factor, the reactivity management has been investigated in details, which has led to the development of a new methodology for optimal used of burnable poisons. The safety analysis against REA highlights new conceptions limits, in particular on the maximal sub-moderating ratio in order to avoid nucleate boiling. It also reveals the consequences on the reactor safety of some design choices such as low soluble boron inventory.

Dans le cadre de la recherche sur les accidents graves pour la deuxième et la troisième génération de réacteurs nucléaires, certains aspects de l'ablation de béton dans le puits de cuve au cours de l'interaction corium-béton (ICB) restent encore inexpliquées. La détermination d'échange de chaleur le long de la région interfaciale entre un bain de corium et un béton est importante pour l'évaluation de la progression d'ablation du béton et, éventuellement, la percée de fondation. Le projet CLARA s'inscrit une recherche expérimentale sur la thermohydraulique au sein d'un bain de liquide agitée par des bulles de gaz. Les essais CLARA sont réalisés avec des matériaux simulants. Ils permettent de mettre en évidence l'influence de la vitesse superficielle du gaz, de la viscosité du liquide et de la géométrie sur le coefficient d'échange de chaleur entre le bain de liquide chauffé et les parois verticales et horizontales de la piscine qui sont maintenues à une température uniforme. La première campagne d'essais a été réalisée avec la configuration du bain de petite taille (50 cm × 25 cm × 25 cm). Les essais ont été réalisés avec des liquides couvrant un large éventail de viscosité dynamique, d'environ 1 mPa s à 10000 mPa s. La vitesse superficielle du gaz est modifiée jusqu'à 8 cm/s. Cette thèse comporte une brève description de la phénoménologie de l'ICB, une synthèse bibliographique sur les corrélations d'échange de chaleur existantes pour l'écoulement diphasique et le taux de vide, une description de l'installation CLARA, les résultats des essais et leur interprétation. Les résultats expérimentaux sont comparés avec les modèles existants et certains nouveaux modèles pour l'évaluation du coefficient d'échange de chaleur dans un écoulement diphasique. / In the context of severe accident research for the second and the third generation of nuclear power plants, there are still open issues concerning some aspects of the concrete cavity ablation during the molten corium - concrete interaction (MCCI). The determination of heat transfer along the interfacial region between the molten corium pool and the ablating basemat concrete is crucial for the assessment of concrete ablation progression and eventually the basemat meltthrough. For the purpose of experimental investigation of thermalhydraulics inside a liquid pool agitated by gas bubbles, the CLARA project has been launched. The CLARA experiments are performed using simulant materials and they reveal the influence of superficial gas velocity, liquid viscosity and pool geometry on the heat transfer coefficient between the internally heated liquid pool and vertical and horizontal pool walls maintained at uniform temperature. The first test campaign has been conducted with the small pool configuration (50 cm × 25 cm × 25 cm). The tests have been performed with liquids covering a wide range of dynamic viscosity from approximately 1 mPa s to 10000 mPa s and the superficial gas velocity is varied up to 8 cm/s. This thesis comprises a brief description of MCCI phenomenology, literature reviews on the existing heat transfer correlations for twophase flow and the void fraction, a description of CLARA setup, experimental results and their interpretation. The experimental results are compared with existing models and some new models for the assessment of heat transfer coefficient in two-phase flow. / U kontekstu istraživanja teških nesre´ca u nuklearnim elektranamadruge i tre´ce generacije, neka pitanja vezana za ablaciju temelja kontejnmentatijekom interakcije rastaljenog korijuma i betona i dalje ostajuotvorena. Odred¯ivanje prijenosa topline u površinskom podrucˇjuizmed¯u bazena rastaljenog korijuma i betona kljucˇno je za odred¯ivanjenapredovanja ablacije i u konaˇcnici procjene vremena rastapanjatemelja kontejnmenta. U svrhu eksperimentalnog istraživanja prijenosatopline u tek´cinama miješanima ubrizgavanjem zraka, pokrenutje projekt nazvan CLARA.CLARA eksperimenti izvode se koriste´ci imitacijske materijale i otkrivajuutjecaj fiktivne brzine plina, viskoznosti teku´cine i geometrijebazena na koeficijent prijenosa topline izmed¯u grijanog bazena te njegovihvetrikalnih i horizontalnih stijenki ˇcija se temperatura održavana konstantnoj temperaturi. Prva serija eksperimenata provedena je sbazenom male konfiguracije (50 cm × 25 cm × 25 cm). Eksperimentisu izvedeni s teku´cinama dinamiˇcke viskoznosti od približno 1 mPas do 10000 mPa s, dok je maksimalna fiktivna brzina plina 8 cm/s.Ova disertacija sadrži kratak opis fenomenologije procesa interakcijerastaljenog korijuma i betona, pregled postoje´cih korelacija zaviprijenos topline u dvofaznom toku i korelacija za poroznost, opisCLARA eksperimentalne postave, rezultate eksperimenta i njihovuinterpretaciju. Rezultati eksperimenta su uspored¯eni s predvid¯anjimaprema postojec´im modelima. Predloženi su takod¯er i neke nove korelacijeza odred¯ivanje koeficijenta prijenosa topline u dvofaznom toku.

McMaster University DOCTOR OF PHILOSOPHY (2022) Hamilton, Ontario (Engineering Physics) TITLE: Development and Evaluation of Polaris CANDU Geometry Modelling and of TRACE_Mac/PARCS_Mac Coupling with RRS for CANDU Analysis AUTHOR: Simon Younan, M.A.Sc. (McMaster University), B.Eng. (McMaster University) SUPERVISOR: Dr. David Novog NUMBER OF PAGES: xiv, 163 / In the field of nuclear safety analysis, as computers have become more powerful, there has been a trend away from low-fidelity models using conservative assumptions, to high-fidelity best-estimate models combined with uncertainty analysis. A number of these tools have been developed in the United States, due to the popularity of light water reactors. These include the SCALE analysis suite developed by ORNL, as well as the PARCS and TRACE tools backed by the USNRC. This work explores adapting the capabilities of these tools to the analysis of CANDU reactors. The Polaris sequence, introduced in SCALE 6.2, was extended in this work to support CANDU geometries and compared to existing SCALE sequences such as TRITON. Emphasis was placed on the Embedded Self-Shielding Method (ESSM), introduced with Polaris. Both Polaris and ESSM were evaluated and found to perform adequately for CANDU geometries. The accuracy of ESSM was found to improve when the precomputed selfshielding factors were updated using a CANDU representation. The PARCS diffusion code and the TRACE system thermalhydraulics code were coupled, using the built-in coupling capability between the two codes. In addition, the Exterior Communications Interface (ECI), used for coupling with TRACE, was utilized. A Python interface to the ECI library was developed in this work and used to couple an RRS model written in Python to the coupled PARCS/TRACE model. A number of code modifications were made to accommodate the required coupling and correct code deficiencies, with the modified versions named PARCS_Mac and TRACE_Mac. The coupled codes were able to simulate multiple transients based on prior studies as well as operational events. The code updates performed in this work may be used for many future studies, particularly for uncertainty propagation through a full set of calculations, from the lattice model to a full coupled system model. / Thesis / Doctor of Philosophy (PhD) / Modern nuclear safety analysis tools offer more accurate predictions for the safety and operation of nuclear reactors, including CANDU reactors. These codes take advantage of modern computer hardware, and also a shift in philosophy from conservative analysis to best estimate plus uncertainty analysis. The goal of this thesis was to adapt a number of modern tools to support CANDU analysis and uncertainty propagation, with a particular emphasis on coupling of multiple interacting models. These tools were then demonstrated, and results analyzed. The simulations performed in this work were successful in producing results comparable to prior studies along with experimental and operational data. This included the simulation of four weeks of reactor operation including “shim mode” operation. Sensitivity and uncertainty analyses were performed over the course of the work to quantify the precision and significance of the results as well as to identify areas of interest for future research.

<p>Among the six GEN-IV reactor concepts recommended by the Gen-IV International Forum, supercritical water-cooled reactors (SCWR) have gained significant interests due to its economic advantage, technology and experience continuity. In the last few years, extensive R&D activities have been launched covering the various aspects of SCWR development, especially in thermal-hydraulic analysis. In Canada, most R&D projects are led by AECL or NRCan.</p> <p>SCWR design and development require the modification of simulation codes used for design and safety demonstration of subcritical water-cooled reactors. This study modifies the subchannel code COBRA-TF, applicable to only subcritical water-cooled reactors, to a new version COBRA-TF-SC, applicable to both supercritical and subcritical water-cooled reactors. Supercritical water property data tables and supercritical water property formulations are implemented. Supercritical water heat transfer and pressure drop correlations are also added. The saturation curve in the subcritical model is extended by introducing a pseudo two-phase region at supercritical pressures to avoid any numerical instabilities consistent with other studies.</p> <p>Some simple fuel bundle experimental data on the flow and temperature distribution are used to evaluate the code. The fuel bundle experiment is simulated with both COBRA-TF-SC and AECL's ASSERT-PV-SC. The COBRA-TF-SC predicted results show good agreement with the experimental data and results obtained from ASSERT-PV-SC, demonstrating good feasibility and accuracy of this code. COBRA-TF-SC is then used to predict the detailed thermalhydraulics behaviour of the 62-element Canadian SCWR fuel bundle design. The advantage of COBRA-TF-SC is that it can accommodate transcritical flow conditions whereas the existing subchannel codes for SCWRs cannot.</p> / Master of Applied Science (MASc)

L’un des accidents de dimensionnement d’un réacteur à eau pressurisée est l’Accident de Perte de Réfrigérant Primaire (APRP). L’évènement initiateur d’un tel accident est une brèche sur le circuit primaire du réacteur entrainant une perte d’inventaire en eau, et de ce fait conduit à un assèchement des assemblages combustibles. En conséquence, une augmentation considérable de la température surviendrait à l’intérieur du cœur du réacteur. Ainsi, les gaines de combustible peuvent éventuellement se déformer et des zones dites ballonnées apparaitre. Ces zones vont avoir un fort impact sur l’efficacité du refroidissement du cœur du réacteur. Pour contribuer à l’étude thermohydraulique d’un APRP, la présente thèse a pour but la caractérisation expérimentale des interactions entre un écoulement diphasique de vapeur/gouttes et une zone partiellement bouchée. Afin de reproduire un tel scénario, le banc expérimental thermohydraulique COLIBRI a été conçu. Plusieurs configurations géométriques de la zone ballonnée, caractéristiques d’un APRP, sont analysées (longueur et taux de bouchage associés au ballonnement). Afin de caractériser les échanges thermiques paroi/fluide ainsi que la dynamique des gouttes, des diagnostics optiques et thermiques sont utilisés : l’Anémométrie Phase Doppler (PDA) pour mesurer le diamètre et la vitesse des gouttes, la Fluorescence Induite par Laser (LIF) pour mesurer la température des gouttes et la Thermographie Infrarouge (IR) afin d’estimer le flux de chaleur extrait du tube par l’écoulement. En parallèle, une modélisation du problème a été développée afin d’obtenir une approche théorique de la capacité de refroidissement de l’écoulement diphasique. Le système d’équations décrivant la conservation de la masse, de la quantité de mouvement et de l’énergie permettra d’estimer l’impact respectif des différents mécanismes de transferts thermiques mis en jeu ainsi que l’évolution spatio-temporelle des paramètres thermohydrauliques / During a Loss of Coolant Accident (LOCA) in a Pressurized Water Reactor (PWR), caused by a break or a leakage on the primary circuit, partial or even complete drying of the fuel assemblies may occur. In these conditions, the fuel temperature increases, leading to a significant deformation and rupture of the fuel rod cladding. The cooling flow might be impaired, according to the size and distribution of the deformed zones within the fuel assemblies during the emergency cooling phase (Reflooding phase). To contribute to the thermalhydraulic study of the reflooding phase, this study aims to characterize experimentally the coolability of a representative deformed sub-channel by a steam-droplets flow under LOCA conditions. In order to reproduce such a scenario, the experimental thermal-hydraulic set-up COLIBRI was designed. Several geometrical blockage configurations are analyzed (Blockage ratios and axial lengths). Three measurement techniques are set up to follow the cooling transient phase of each experience: Phase Doppler Anemometry (PDA) in order to obtain both velocity and diameter of droplets, Laser Induced Fluorescence (LIF) to measure the mean droplet temperature and Infrared thermography to estimate the heat flux removed by the two-phase flow. Additionally, a one-dimensional mechanistic model, taking into account of the heat transfers mechanisms in the post-dry out region, is developed in order to analyze the experimental data and identify each one of the wall-to-fluid heat transfers (radiation with vapor and droplets, convection with vapor, evaporation, and droplet impact)