Nuclear fusion is considered a clean energy source: it emits no CO2 and leaves little radioactive waste. It is important to start paving the path toward nuclear fusion whilst simultaneously moving away from fossil fuels and carbon emissions. One of the challenges of nuclear fusion is the lack of tritium, which, together with deuterium makes up its fuel. This research is focused on utilizing one current method of nuclear fission technology, namely molten salt reactors, to generate at least the initial loads of tritium for the first fusion reactors. Current research is primarily focused on providing tritium during the nuclear fusion reaction. However, it is also necessary to have a tritium supply whenever we start up a nuclear fusion reactor. The largest source of tritium is the CANDU nuclear fission reactor. A typical 500 MW CANDU produces 130 g of tritium annually as a biproduct of power generation. However, a future commercial fusion power plant is expected to consume 300 g of tritium per day to produce 800 MW. Thus, this research explores the possibility of breeding tritium in other fission reactors, in particular molten salt reactors (MSR). MCNP4C was used to simulate a simple Molten Salt Reactor setting with 61 molten salt fuel channels and applying a molten salt blanket to study how the presence of specific elements in the blanket affects tritium production, as well as criticality. The study relies on nuclear data from the National Nuclear Data Center (NNDC), and Oak Ridge National Laboratory (ORNL) as benchmark to verify the accuracy of the results. The calculated output of tritium is 325 g/year for a 100 MW (th) reactor, which is considered a positive outcome that opens the door for more research in this direction. / Thesis / Master of Applied Science (MASc)

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Master’s thesis focuses on the history and current progress in research of molten salt reactors around the world, with an emphasis placed on the properties of molten salts and the problems associated with their use. In relation to the practical part, one chapter is devoted to the creation of input file in the MCNP software. The practical part deals with neutron activation analysis of graphite prism experiment, which is filled with powder NaCl salt. This experiment is focused on the effect of salt on neutron flux with different energy levels. The whole problem was also simulated in the MCNP environment along with the experiment. At the end of the thesis, the individual methods are compared and evaluated.

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Corrosion is a major challenge in the use of molten fluoride salt as a coolant in molten salt reactors (MSRs). A promising way of satisfying the two requirements of high strength and corrosion resistance is to clad structural alloys with a corrosion resistant material. Four candidate cladding and structural alloy combinations—stainless steel 316L and Incoloy 800H structural alloys either diffusion bonded to Hastelloy N or electroplated with nickel—were thermally aged at 700 °C for two to eight months. Based on measured concentration profles, the diffusion resistance of the four material combinations was compared and diffusion results were extrapolated to an end of reactor lifetime. The most important conclusion from this work is that Hastelloy N is highly likely to be signifcantly more diffusion resistant than nickel. The difference in diffusion resistance between Incoloy 800H and stainless steel 316L is relatively small. Two methods were used for extrapolating experimental diffusion results: (1) a diffusion model and calculated diffusion coeffcients and (2) simulations with Thermo-Calc DICTRA. Some simulations were carried out with a corrosion boundary condition of near-zero chromium concentration, demonstrating the potential of simulations for predicting diffusionlimited corrosion in molten fluoride salts. A surprising result of these simulations is that decreasing the thickness of Ni plating did not increase the thickness of diffusion zones in underlying structural alloys. / Thesis / Master of Applied Science (MASc)

The first Molten Salt Reactor (MSR) was designed and tested at Oak Ridge National Laboratory (ORNL) in the 1960's, but recent technological advancements now allow for new components, such as heat exchangers, to be created for the next generation of MSR's and molten salt-cooled reactors. The primary (fuel salt-to-secondary salt) heat exchanger (PHX) design is shown here to make dramatic improvements over traditional shell-and-tube heat exchangers when changed to a compact heat exchanger design. While this paper focuses on the application of compact heat exchangers on a Molten Salt Reactor, many of the analyses and results are similarly applicable to other fluid-to-fluid heat xchangers. The heat exchanger design in this study seeks to find a middle-ground between shell- and-tube designs and new ultra-efficient, ultra-compact designs. Complex channel geometries and microscale dimensions in modern compact heat exchangers do not allow routine maintenance to be performed by standard procedures, so extended surfaces will be omitted and hydraulic diameters will be kept in the minichannel regime (minimum channel dimension between 200 μm and 3 mm) to allow for high-frequency eddy current inspection methods to be developed. High aspect ratio rectangular channel cross-sections are used. Various plant layouts of smaller heat exchanger banks in a "modular" design are introduced. FLUENT was used within ANSYS Workbench to find optimized heat transfer and hydrodynamic performance. With similar boundary conditions to ORNL's Molten Salt Breeder Reactor's shell-and-tube design, the compact heat exchanger interest in this thesis will lessen volume requirements, lower fuel salt volume, and decrease material usage. / Master of Science

This is a report of a verification study of the multiphysics simulation framework MOOSE which was preformed at the company Seaborg Technologies. In the process of designing molten salt reactors there is a special need of making credible multiphysics simulations since the fuel is in motion. In this study the incompressible version of Navier-Stokes equations of finite volumes available in the Navier-Stokes module of the MOOSE framework is verified by modelling and simulations of fluid flow and heat transfer in two different systems with available benchmarks. The first system, a thin buoyancy driven molten sodium hydroxide test loop which is verified by a similar model made with the high fidelity CFD software STAR-CCM+ as benchmark. The second system, forced convection of air through a straight pipe with heated walls which is verified by comparisons with an analytical solution. The resulting velocity profiles from simulations of the first system corresponds well with the benchmark but certain conclusions can not be drawn from it since the the transient simulations stops to converge before reaching equilibrium. The results from simulations of the second system corresponds well with the analytical solution and no convergence issues arise. The conclusion from the results is that the incompressible version of Navier-Stokes equations of finite volumes available in the Navier-Stokes module of the MOOSE framework has potential to be used in multiphysics simulations of molten salt reactors but seemingly not in cases of buoyancy driven flows in thin geometries. Two proposals for further work is recommended. The first is that this implementation is applied in a context with forced fluid flow or a context with thicker fluid domain. The second proposal is that the other available abilities of MOOSE such as finite element method and/or the compressible version of the Navier-Stokes equations should be tested.

Un concept innovant de réacteurs nucléaires à sels fondus, le Thorium Molten Salt Reactor (TMSR), a été défini au LPSC Grenoble. Le présent mémoire porte sur les études, optimisations et caractérisations réalisées sur les configurations en spectre rapide de ce concept, appelées ‘TMSR non modérés' ou TMSR-NM, très prometteuses. Le cœur est un simple cylindre dans lequel circule un sel fluore contenant du LiF et le combustible. Nos études portent sur les caractéristiques de ces réacteurs en termes de sûreté, inventaire fissile, retraitement chimique, production de déchets et capacité de régénération et de déploiement. Un tel réacteur présente maints avantages intrinsèques permettant un fonctionnement simple et sûr en cycle du combustible Thorium, ainsi que l'utilisation de divers éléments fissiles au démarrage tels l'233U, 235U ou les transuraniens issus des réacteurs actuels. Ceci permettrait une transition optimisée vers le cycle Thorium tout en fermant le cycle actuel.

Les réacteurs nucléaires de 4ème génération devront permettre une utilisation optimisée desressources naturelles. Les travaux réalisés durant cette thèse se placent ainsi dans le cadre del’étude du potentiel de déploiement d’un tel réacteur : le MSFR (Molten Salt Fast Reactor), réacteurà sels fondus à spectre neutronique rapide dans une configuration innovante et encore peuétudiée. Comme un excellent niveau de sûreté est une condition nécessaire pour le déploiementde l’énergie nucléaire, il est important de soulever la question de la sûreté de ce type de réacteurdès les premières phases de sa conception.Le MSFR a fait l’objet d’études comparatives des outils de simulations numériques dans lecadre d’un benchmark neutronique au sein du projet européen EVOL. La définition et l’analysedu benchmark neutronique statique et en évolution ont été réalisées pendant cette thèse. Lescomparaisons des différentes grandeurs physiques ont permis de conclure à un bon accord entreles différents codes et méthodes utilisés par les partenaires du projet, et ont mis en avant l’influencedu choix des bases de données nucléaires. Dans l’objectif de l’étude de sûreté du MSFR,la puissance résiduelle a aussi été étudiée en détails. Un outil de calcul de chaleur résiduellea été développé et validé, permettant ainsi d’évaluer la puissance résiduelle précise du MSFR.Les sources de chaleur de chaque localisation contenant des produits radioactifs ont alors étéquantifiées. Ceci a permis de conclure que le sel combustible et l’unité de bullage constituent lessources majeures de puissance résiduelle.Nous avons initié un travail sur la méthodologie de l’étude de sûreté. Les principes fondamentauxde sûreté sont directement transposables au MSFR, mais leurs applications concrètes nele sont pas. En effet, la spécificité du design, due à l’état liquide du combustible et aux systèmesde retraitement associés au réacteur, ainsi que l’état embryonnaire du design, font qu’un travailpréliminaire de transposition des éléments de sûreté a dû être réalisé. Ce travail a conduit entreautres à dresser une liste d’accidents propres au MSFR. Enfin, nous avons pu mener des étudesphysiques préliminaires sur les conséquences possibles de certains de ces accidents, qui serontutilisées comme base pour des études plus approfondies avec des outils plus sophistiqués. / The nuclear reactors of the 4th generation must allow an optimized use of natural resources,while performing at a high safety level. The framework of this thesis is the deployment study ofone of such a system, an innovative and still little studied Molten Salt Fast Reactor. An excellentsafety is an ultimate requirement of the nuclear energy deployment, so it is important to raisethis question at the current early stage of the MSFR concept development.This concept was the subject of a neutronic tool benchmark within a European projectEVOL. Definition, calculations and results analyses were performed during this thesis. Comparisonsof static neutronic and burn-up calculations, performed by the project participants,concluded to a good agreement between the different codes and methods used and pointed outthe sensibility of the nuclear database choice on the results. With the aim of safety analysis ofthe MSFR, the decay heat was studied in detail. The tool used for the decay heat calculationwas developed and validated, to finally evaluate the decay heat in the reactor. The decay heatsource presented in different zones was quantified, concluding to a high importance of the coolingof the fuel salt and the bubbling system enclosing a part of the fission products.The safety analysis methodology was also studied in this thesis. Even if the safety principlesare directly transposable to the MSFR, the precise recommendations are not. This is due to thespecificity of the design that relies on the liquid state of the fuel, on the reprocessing systemslocated in the reactor and the embryonic stage of the design. First, a preliminary transpositionwork of some criteria to the MSFR design was realized, resulting amongst other things in a listof accidental scenarios particular for MSFR. Finally, a preliminary physical study of some typesof accidental scenarios was performed, that can be used as a basis for further analyses with moresophisticated tools.