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.

The Swedish Nuclear Fuel and Waste Management Company (SKB) has developed a method for final disposal of spent nuclear fuel. This technique requires accurate measurement of the residual decay heat of every assembly. For this purpose, depletion codes as well as calorimetric and gamma-ray spectroscopy experimental methods have been developed and evaluated. In this work a prototype analysis tool has been developed to automate the analysis of both calorimetric and gamma-ray spectroscopy measurements. The performance of the analysis tool has been investigated by comparing its output with earlier results and calculations. Parallel to the software development, new measurements on 73 BWR assemblies were performed. The results obtained for the determination of the residual decay heat are presented. Finally, suggestions for further development are outlined and discussed.

This master thesis has been conducted at Forsmark nuclear power plant. In nuclear power plants, safety is one of the most important tasks. After the Fukushima-accident, Forsmarks Kraftgrupp AB investigated the possibility to have additional emergency cooling systems. This thesis aims to investigate the use of cooling towers as mobile emergency cooling system for decay heat removal for the nuclear reactors at Forsmark NPP. The advantages and disadvantages of different cooling tower types were looked at. To achieve mobility, some restraints were considered such as dimensions and weight. In the design of cooling towers, experimental values are used. As these crucial data are missing, a backfitting from actual designs was done to derivate the missing data. The results of this thesis show that wet cooling towers are in this case more effective than dry cooling towers. The wet cooling towers cover in the best scenario up to 38MW which is 10 times higher than what can be achieved with the dry cooling tower. That covers the maximal decay heat released in the reactor F3 12 hours after a shutdown.

It is important to know the heat generated due to nuclear decay in the final repository for spent nuclear fuel. In Sweden, the heating powers generated in spent nuclear fuels are currently measured in the calorimeter System 251 at the Clab facility, Oskarshamn. In order to better measure, and increase understanding, of the temperature measurements in the calorimeter, a simple thermodynamic model of its temperature evolution was developed. The model was described as a system of ordinary differential equations, which were solved, and the solution was applied to calibration measurements of the calorimeter. How precise the model is, how its parameters affect the model, et cetera, are addressed. How the temperature evolution of the system changes as the values of parameters in the model are changed is addressed. The mass correction of the calorimeter could be estimated from this model, which validated the established mass correction of the calorimeter. How the measurement results from the calorimeter would be affected if the volume of the calorimeter was changed was also considered. Additionally, gamma radiation escape from the calorimeter without being detected as heat in the calorimeter. The gamma escape energy fraction was estimated by SERPENT simulations of the calorimeter, as a function of the initial photon energy. The gamma escape was also estimated for different values of the radius of System 251.