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.

Inlet headers in the primary heat transport system(PHTS) of CANDU type reactors, are used to collect the coolant coming from the steam generators and distribute them into the reactor core via several feeders. During a postulated loss of coolant accident (LOCA), depressurization and vapor supplement into the core may occur, which results a deterioration in the heat transfer from fuel to the coolant. When a depressurization occurs, &ldquo / Emergency Coolant Injection(ECI)&rdquo / system in the PHTS in CANDU reactors, is automatically become active and supply coolant is fed into the reactor core via the inlet header and feeders. . This study is focused on the experimental and computational investigation of the ECI effect during a LOCA in a CANDU inlet header. The experiments were carried out in METU Two-Phase Flow Test Facility which consists of a scaled CANDU inlet header having 5 connected feeders. The same tests were simulated with a one dimensional two-fluid computer code, CATHENA, developed by Atomic Energy of Canada Limited(AECL). The average void fraction and the two phase mass flowrate data measured in the experiments are compared with the results obtained from CATHENA simulation. Although a few mismatched points exist, the results coming from two different studies are mostly matching reasonably. Lack of three-dimensional modeling for headers in CATHENA and experimental errors are thought to be the reasons for these dismatches.

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.

In the field of nuclear medicine, radioisotopes are used for applications such as diagnostic imag- ing, treatment, and equipment sterilization. The most commonly used radioisotope in medicine is technetium-99m (Tc-99m). It is used in 80% of all nuclear medicine procedures. Its parent isotope is molybdenum-99 (Mo-99). NRU, which is now closed, formerly produced 40% of the worlds demand for Mo-99. The production capacity of this reactor has been supplemented by a network of cyclotrons and a modified research reactor. This study aims to provide an alternative means of production for Mo-99, as well as other radioisotopes by modifying the center pin of a standard 37-element bundle of a CANDU reactor. The neutron transport code DRAGON, and the neutron diffusion code DONJON were used to model a CANDU-9 reactor. The lowest, median, and highest power channels were chosen as candi- dates for the modified bundles. It was found that the reactor parameters were altered by a negligible amount when any one channel was used to house the modified bundles. Significant quantities of the radioisotope lutetium-177 as well as the generating isotopes of the alpha-emitting radioisotopes lead- 212/bismuth-212, and radium-223 were produced. However, only minute amounts of molybdenum-99, and the generating isotope of bismuth-213 were produced. / Thesis / Master of Applied Science (MASc)

In CANDU reactors, under normal operating conditions, the inlet headers collect and distribute single-phase liquid flow (heavy water) to the fuel cooling channels via the feeders. However, under some postulated loss of coolant accidents, the inlet headers may receive two-phase fluid (steam/water) and the fluid forms a stratified region inside the header. To predict the thermalhydraulic behaviour of headers for the reactor safety analysis, the two-phase flow distribution within the headers and through the feeders must be modelled. In order to analyse the two-phase flow behaviour of a scaled CANDU inlet header / a transparent and instrumented version of a header with 5 feeders was previously built in the Mechanical Engineering Department of Middle East Technical University (METU-Two Phase Flow Test Facility / METU-TPFTF). The aim of this study is to investigate two-phase flow distribution through multiple outlets from such a horizontal drum both numerically and experimentally. For this purpose, three-dimensional incompressible finite difference equations in cylindrical coordinates were derived by using two-fluid model to simulate adiabatic two-phase flow (air/water) in the header numerically. The discretized equations were then programmed into a computer code which was developed specifically for modelling the header type geometry. A method based on the principles of Implicit Multi Field (IMF) technique has been utilised to solve those equations. The solution algorithm was tested by using some numerical benchmark problems. A number of experimental tests covering single and two-phase flow distribution through outlet pairs from the header were performed. Void fractions and flow rates obtained from these tests are in good agreement with the code results. The code also predicts the void fraction and pressure distribution in the header satisfactorily.

Probabilistic assessment and life cycle management of engineering components and systems in a nuclear power plant is intended to ensure safe and efficient operation of energy generation over its entire life. The CANDU reactor core consists of 380-480 pressure tubes, which are like miniature pressure vessels that contain natural uranium fuel. Pressure tubes operate under severe temperature and radiation conditions, which result in degradation with ageing. Presence of flaws in a pressure tube makes it vulnerable to delayed hydride cracking (DHC), which may lead to rupture or break-before-leak situation. Therefore, assessment of flaws in the pressure tubes is considered an integral part of a reactor core assessment program. The main objective of the thesis is to develop advanced probabilistic and mechanical stress field models for the assessment of flaws. The flaw assessment models used by the industries are based on deterministic upper/lower bound values for the variables and they ignore uncertainties associated with system parameters. In this thesis, explicit limit state equations are formulated and first order reliability method is employed for reliability computation, which is more efficient than simulation-based methods. A semi-probabilistic approach is adopted to develop an assessment model, which consists of a mechanics-based condition (or equation) involving partial factors that are calibrated to a specified reliability level. This approach is applied to develop models for DHC initiation and leak-before-break assessments. A novel feature of the proposed method is that it bridges the gap between a simple deterministic analysis and complex simulations, and it is amenable to practical applications. The nuclear power plant systems are not easily accessible for inspection and data collection due to exposure to high radiation. For this reason, small samples of pressure tubes are inspected at periodic intervals and small sample of data so collected are used as input to probabilistic analysis. The pressure tube flaw assessment is therefore confounded by large sampling uncertainties. Therefore, determination of adequate sample size is an important issue. In this thesis, a risk informed approach is proposed to define sample size requirement for flaw assessment. Notch-tip stress field is a key factor in any flaw assessment model. Traditionally, linear elastic fracture mechanics (LEFM) and its extension, serves the basis for determination of notch-tip stress field for elastic and elastic-perfectly-plastic material, respectively. However, the LEFM solution is based on small deformation theory and fixed crack geometry, which leads to singular stress and strain field at the crack-tip. The thesis presents new models for notch and crack induced stress fields based on the deformed geometry. In contrast with the classical solution based on small deformation theory, the proposed model uses the Cauchy's stress definition and boundary conditions which are coupled with the deformed geometry. This formulation also incorporates the rotation near the crack-tip, which leads to blunting and displacement of the crack-tip. The solution obtained based on the final deformed configuration yields a non-singular stress field at the crack-tip and a non-linear variation of stress concentration factor for both elastic and elastic-perfectly-plastic material. The proposed stress field formulation approach is applied to formulate an analytical model for estimating the threshold stress intensity factor (KIH) for DHC initiation. The analytical approach provides a relationship between KIH and temperature that is consistent with experimental results.

Probabilistic assessment and life cycle management of engineering components and systems in a nuclear power plant is intended to ensure safe and efficient operation of energy generation over its entire life. The CANDU reactor core consists of 380-480 pressure tubes, which are like miniature pressure vessels that contain natural uranium fuel. Pressure tubes operate under severe temperature and radiation conditions, which result in degradation with ageing. Presence of flaws in a pressure tube makes it vulnerable to delayed hydride cracking (DHC), which may lead to rupture or break-before-leak situation. Therefore, assessment of flaws in the pressure tubes is considered an integral part of a reactor core assessment program. The main objective of the thesis is to develop advanced probabilistic and mechanical stress field models for the assessment of flaws. The flaw assessment models used by the industries are based on deterministic upper/lower bound values for the variables and they ignore uncertainties associated with system parameters. In this thesis, explicit limit state equations are formulated and first order reliability method is employed for reliability computation, which is more efficient than simulation-based methods. A semi-probabilistic approach is adopted to develop an assessment model, which consists of a mechanics-based condition (or equation) involving partial factors that are calibrated to a specified reliability level. This approach is applied to develop models for DHC initiation and leak-before-break assessments. A novel feature of the proposed method is that it bridges the gap between a simple deterministic analysis and complex simulations, and it is amenable to practical applications. The nuclear power plant systems are not easily accessible for inspection and data collection due to exposure to high radiation. For this reason, small samples of pressure tubes are inspected at periodic intervals and small sample of data so collected are used as input to probabilistic analysis. The pressure tube flaw assessment is therefore confounded by large sampling uncertainties. Therefore, determination of adequate sample size is an important issue. In this thesis, a risk informed approach is proposed to define sample size requirement for flaw assessment. Notch-tip stress field is a key factor in any flaw assessment model. Traditionally, linear elastic fracture mechanics (LEFM) and its extension, serves the basis for determination of notch-tip stress field for elastic and elastic-perfectly-plastic material, respectively. However, the LEFM solution is based on small deformation theory and fixed crack geometry, which leads to singular stress and strain field at the crack-tip. The thesis presents new models for notch and crack induced stress fields based on the deformed geometry. In contrast with the classical solution based on small deformation theory, the proposed model uses the Cauchy's stress definition and boundary conditions which are coupled with the deformed geometry. This formulation also incorporates the rotation near the crack-tip, which leads to blunting and displacement of the crack-tip. The solution obtained based on the final deformed configuration yields a non-singular stress field at the crack-tip and a non-linear variation of stress concentration factor for both elastic and elastic-perfectly-plastic material. The proposed stress field formulation approach is applied to formulate an analytical model for estimating the threshold stress intensity factor (KIH) for DHC initiation. The analytical approach provides a relationship between KIH and temperature that is consistent with experimental results.

Le niveau de conversion des réacteurs CANDU et REP en cycle thorium a été étudié dans l'optique d'une utilisation en troisième et dernière strate de scénarios symbiotiques. Le plutonium du combustible REP usé serait par exemple utilisé en CANDU Th/Pu pour produire de l'233U, qui alimenterait ces réacteurs à eau et haute conversion. En cas d'augmentation importante de la production d'énergie à partir d'uranium, cette alternative basée sur des réacteurs existants pourrait suppléer une IVe génération trop tardive. Pour évaluer la compétitivité de tels scénarios, des calculs de cycles détaillés ont été effectués selon une méthodologie de simulation de coeur développée pour le CANDU-6 et adaptée au REP de type N4. Le CANDU Th/233U enrichi à 1.30 wt% est régénérateur, avec un burnup court de 7 GWj/t. Augmenter légèrement l'enrichissement allonge considérablement le cycle, au prix d'une sous-génération. Multirecycler conduit également à une perte de conversion, qui peut néanmoins être compensée par un chargement fissile hétérogène. La conversion à puissance standard est moins bonne en REP Th/233U qu'en CANDU (inventaire fissile réduit de moitié après 50 GWj/t) mais peut être améliorée par sous-modération. L'analyse neutronique montre que l'essentiel du gap de conversion entre CANDU et REP vient des conditions opératoires économes en neutrons du CANDU. Des scénarios ont été comparés du point de vue de l'économie d'uranium et de l'aval du cycle dans les deux cas, et ont confi rmé l'intérêt du CANDU. Deux pistes de recherche ont été identi fiées : l'évaluation de la sûreté des CANDUs au thorium par cinétique avec contre-réactions thermiques, et l'étude de coeurs fortement sous-modérés en cuve standard de REP.