This master’s thesis deals with a design of Na-CO2 heat exchanger. There is a comparison of shell and tube heat exchanger with PCHE in the first part. Further the shell and tube heat exchanger with internally finned tubes was chosen as the basic conception. There was performed an optimization of construct and operations parameters for this concept. The optimization was performed on the basis of thermal and hydraulic calculations. Further there were performed calculations for ensuring of safe operation of the heat exchanger. The conclusion of the diploma thesis deals with comparison of the designed heat exchanger with similar projects.

Molten Salt Fast Reactors (MSFRs) have prominent advantages such as fuel breeding, nuclear waste transmutation, and inherent safety. They are the only liquid-fueled nuclear reactors currently receiving significant attention as fourth-generation advanced nuclear systems. To address the challenges of short operational lifetimes and proliferation issues during online fuel processing, the breed-and-burn (B&B) MSFR is among the most promising advanced reactor types. In this study, a large-volume B&B MSFR model without online actinide element treatment was simulated and analyzed using the Monte Carlo simulation software SERPENT, considering different power levels and sizes of the inactive core. The results demonstrate that, under otherwise identical conditions, the operational lifetime and conversion ratio of MSFRs increase with decreasing power levels and increasing the size of the inactive core. These findings provide a foundation and theoretical basis for achieving B&B MSFRs without online actinide element treatment. / Smält saltsnabbreaktorer (MSFRs) har framträdande fördelar såsom bränsleförädling, transmutation av kärnavfall och inneboende säkerhet. De är de enda flytande drivna kärnreaktorerna som för närvarande får betydande uppmärksamhet som fjärde generationens avancerade kärnkraftverk. För att möta utmaningarna med korta driftstider och spridningsproblem vid online bränslebearbetning är rask-och-bränning (B&B) MSFR bland de mest lovande avancerade reaktortyperna. I denna studie simulerades och analyserades en storskalig B&B MSFR-modell utan behandling av aktinidelement online med hjälp av Monte Carlo simuleringsprogramvaran SERPENT, med hänsyn till olika effektnivåer och storlekar på den inaktiva kärnan. Resultaten visar att livslängden och konverteringsförhållandet för MSFRs ökar under annars identiska förhållanden i takt med att effektnivåerna minskar och storleken på den inaktiva kärnan ökar. Dessa resultat ger en grund och teoretisk grund för att uppnå B&B MSFRs utan behandling av aktinidelement online.

With the initiative taken by the SUNRISE project (Sustainable Nuclear Energy Research in Sweden) to construct a Lead-cooled research reactor, this thesis intends to extend the knowledge within nuclear fuel development. By using integral iterative modelling and simulating techniques that mimic real-world phenomena, novel fuel materials like uranium nitride are assessed for future validation.  The work deals with the fuel performance analysis of the SUNRISE LFR, employing the TRANSURANUS fuel performance code. This code contains a collection of model parameters that simulate the thermo-mechanical behaviour of the fuel cladding system on an engineering scale of the reactor core. A comparative study is performed for UO$_2$ and UN fuels using the same input data such as fuel geometry. In addition, predefined information relating to the neutronics analysis for the reactor was used as input to the TRANSURANUS code along with literature reviews to select the accurate models on the reactor, fuel, and its behaviour. Furthermore, a sensitivity study is carried out to assess the models and parameters affected by more significant uncertainty.  The uncertainty analysis of the UN fuel's swelling models is performed using the Dakota toolkit. The sampling of input data using the Dakota software coupled with the nuclear simulation program TRANSURANUS produced partial rank correlation coefficients significant to the modelling. However, since the assessed models displayed the same correlation coefficients, the results conclude that a deeper understanding of the theoretical swelling model might be required. / I samverkan med initiativet av SUNRISEprojektet (Sustainable Nuclear Energy Research inSweden) som syftar att bygga en blykyld forskningsreaktor, avser denna avhandling att utökakunskapen inom kärnbränsleutveckling. Med användning av integral iterativ modellering ochsimuleringstekniker som efterliknar verkliga fenomen bedöms nya bränslematerial somuranmononitrid för framtida validering. Arbetet behandlar analysen av bränsleprestanda för SUNRISE LFR, med användning avTRANSURANUS bränsleprestandakod. Denna kod innehåller en samling modellparametrarsom simulerar det termomekaniska beteendet hos bränslebetäckningssystemet i en tekniskskala för reaktorkärnan. En jämförande studie utförs för UO2 och UN-bränslen med sammaingångsdata som t.ex bränslegeometrin. Dessutom användes fördefinierad information om denneutroniska analysen för reaktorn som ingångsdata till TRANSURANUSkoden tillsammans medgranskning av litteratur för att välja lämpliga modeller för reaktorn, bränslet och dess beteende.Därtill genomfördes en känslighetsstudie för att bedöma de modeller och parametrar sompåverkas av mer betydande osäkerhet. Osäkerhetsanalysen av UN-bränslets svällningsmodeller utförs med hjälp av Dakota-verktyget.Samlingen av indata med Dakota-programmet i kombination medkärnkraftssimuleringsprogrammet TRANSURANUS gav korrelationskoefficienter för partiell rangviktiga för modelleringen. Eftersom de utvärderade modellerna visade sammakorrelationskoefficienter, tyder slutsatsen på att en djupare förståelse av den teoretiskasvällningsmodellen krävs

La démarche de conception et de sûreté du réacteur ASTRID, démonstrateur de Réacteur à Neutrons Rapides refroidi au Sodium, implique la modélisation de scénarios d’accidents graves qui font intervenir une fusion du cœur du réacteur. L’objectif de la thèse, en soutien à cette modélisation, est de contribuer à l’identification des processus susceptibles de faire bifurquer un scénario d’accident grave. Deux phases d’un scénario sont traitées pour cela. Tout d’abord, le début d’une séquence de perte de débit primaire non protégée est analysé à l’aide d’un critère analytique développé pendant la thèse, visant à prédire la bifurcation de la décroissance du débit vers un état stabilisé ou bien vers un état instable, menant à la dégradation du cœur. Ce nouveau critère, qui présente l’intérêt de tenir compte de l’effet de l’évolution de la puissance sur la stabilité du débit, est vérifié à l’aide d’un outil de calcul dédié aux accidents de perte de débit non protégés. Dans un second temps, les processus prépondérants impliqués dans une vaporisation de combustible liquide suivie d’une détente de sa vapeur, consécutives à une excursion de puissance accidentelle, sont identifiés via une analyse dimensionnelle. En reprenant les résultats de cette analyse, un outil de calcul est par la suite développé, dont l’objet est de déterminer l’énergie mécanique transmise à la cuve du réacteur lors de la détente. La question du transfert thermique entre la vapeur de combustible se détendant et le caloporteur est particulièrement étudiée. Cet outil est validé via une comparaison à des résultats expérimentaux et à des résultats de calculs issus d’un autre code. Des études paramétriques permettent enfin de quantifier la variabilité des résultats due au choix de modélisation et aux incertitudes sur les données physiques employées. / Severe accidents’ modeling is required for the design and safety analysis of ASTRID, a Generation IV Sodium-cooled Fast Reactor under development in France. This thesis aims at contributing to identify the driving processes of ASTRID’s severe accidents scenarios. First, a stability criterion is developed to analyze the beginning of an unprotected loss of flow accident. This stability criterion assesses whether the decreasing flow is stable or unstable, leading to the core disassembly. This criterion also considers power variations during the loss of flow, which former stability criteria do not take into account. Then, the driving processes of a transient involving a fuel vaporisation followed by its vapor expansion are identified using a dimensional analysis. The simplifications justified by this dimensional analysis are considered further to develop a numerical tool that computes the mechanical energy transmitted to the core vessel in case of fuel vaporisation. The thermal exchange between the expanding fuel vapor and the sodium coolant is especially analyzed. The tool is validated by comparing its results to experimental measures and to another tool’s computations. In the end, parametric studies are done in order to assess the tool computations’ variability induced by physical uncertainties or modeling options.

Contrairement à ses prédécesseurs (Phénix, Super-Phénix, EFR…), le réacteur à neutrons rapides refroidi au sodium (RNR-Na) de IVième génération doit justifier un niveau de sûreté élevé tout en étant à la fois viable économiquement et non-proliférant. Profitant d’un large retour d’expérience, les combustibles de type Oxyde (U,Pu)O2 représentent actuellement la solution de référence en France. Cependant, les combustibles de type carbure (U,Pu)C sont considérés comme une option innovante pour apporter à la conception d’un RNR-Na des degrés d’optimisation supplémentaires. L’objectif de cette thèse était donc de mettre en avant les potentialités du combustible carbure en concevant un cœur de RNR-Na à la fois attractif d’un point de vue économique et au comportement naturel en transitoire incidentel. Pour un parc de réacteurs français, on s’intéressera plus particulièrement à des cœurs iso-générateurs de forte puissance électrique (1500 MWe).Cet objectif a requis la mise en place d’une approche pluridisciplinaire prenant en compte les contraintes de thermomécanique combustible et de thermo-hydraulique en transitoire incidentel dès les premières étapes de la conception. Des modèles simplifiés basés sur les contre-réactions globales K, G et H ont été développés pour estimer le comportement d’un projet de cœur en transitoire de type insertion de réactivité, perte de débit primaire et/ou secondaire. L’avantage de cette nouvelle approche est surtout d’apporter au concepteur des outils complémentaires l’aidant à avoir une vision globale des problématiques de conception, mettant ainsi en avant les innovations ou les paramètres à optimiser pour améliorer les performances d’un cœur de RNR-Na.Cette approche a été appliquée à la conception de cœurs à combustibles carbure avec des performances très intéressantes. Un cœur de forte puissance électrique est proposé : il est isogénérateur de faible volume, avec un inventaire fissile initial réduit et un comportement naturel en transitoire incidentel très satisfaisant. Cependant, le taux de combustion d’une aiguille carbure dans un tel cœur semble limité à 100 MWj/kg à cause du gonflement important du carbure et de sa faible capacité à fluer, ce qui conduit rapidement à l’Interaction Mécanique Pastille-Gaine. Une aiguille fonctionnant à forte puissance linéique nécessite à la fois un jeu pastille-gaine épais et un joint sodium pour retarder l’IMPG, mais aussi un acier de gainage capable d’accommoder l’interaction par son fluage.Les performances en irradiation d’un combustible carbure pour un cœur industriel semblent donc très inférieures à celles obtenues expérimentalement dans le FBTR, où des aiguilles ont atteint un taux de combustion maximal de 155 MWj/kg. Cette différence a été étudiée et en partie expliquée, notamment par la fluence beaucoup plus faible obtenue dans un réacteur expérimental, retardant le critère de gonflement volumique. Deux voies d’exploration ont été mises en évidence pour augmenter les performances du carbure tel qu’utilisé dans un réacteur industriel. La première utilise un jeu pastille-gaine avec une technologie de type « buffer » pour retarder l’IMPG. La seconde est un cœur de faible fluence utilisant un enrichissement accru en plutonium. Les résultats préliminaires obtenus montrent que des taux de combustion supérieurs à 100 MWj/kg devraient être atteignables.Pour conclure, l’approche de conception pluridisciplinaire mise en place au cours de cette thèse s’est révélée efficace pour mettre en avant les avantages du combustible de type carbure. Celle-ci a permis de concevoir une image de cœur de RNR-Na attractive d’un point de vue économique, avec un comportement pardonnant en transitoire accidentel et capable d’atteindre un taux de combustion élevé. / Compared with earlier plant designs (Phénix, Super-Phénix, EFR), GEN IV Sodium-cooled Fast Reactor requires improved economics while meeting safety and non-proliferation criteria. Mixed Oxide (U-Pu)O2 fuels are considered as the reference fuels due to their important and satisfactory feedback experience. However, innovative carbide (U-Pu)C fuels can be considered as serious competitors for a prospective SFR fleet since carbide-fueled SFRs can offer another type of optimization which might overtake on some aspects the oxide fuel technology. The goal of this thesis is to reveal the potentials of carbide by designing an optimum carbide-fueled SFR with competitive features and a naturally safe behavior during transients. For a French nuclear fleet, a 1500 MW(e) break-even core is considered.To do so, a multi-physic approach was developed taking into account neutronics, fuel thermo-mechanics and thermal-hydraulic at a pre-design stage. Simplified modeling with the calculation of global neutronic feedback coefficients and a quasi-static evaluation was developed to estimate the behavior of a core during overpower transients, loss of flow and/or loss of heat removal transients. The breakthrough of this approach is to provide the designer with an overall view of the iterative process, emphasizing the well-suited innovations and the most efficient directions that can improve the SFR design project.This methodology was used to design a core that benefits from the favorable features of carbide fuels. The core developed is a large carbide-fueled SFR with high power density, low fissile inventory, break-even capability and forgiving behaviors during the unscrammed transients studied that should prevent using expensive mitigate systems. However, the core-peak burnup is unlikely to significantly exceed 100 MWd/kg because of the large swelling of the carbide fuel leading to quick pellet-clad mechanical interaction and the low creep capacity of carbide. Moderate linear power fuel pins require both a large initial sodium-bonded gap, delaying the fuel clad mechanical interaction, and a clad able to accommodate it by its high irradiation creep capacity.Irradiated carbide fuel performances predicted for an industrial SFR design are lower than the one obtained in the FBTR reactor irradiations, where 155 MWd/kg was obtained. This difference was studied and partly explained by the lower flux of experimental reactor delaying the embrittlement criterion. Innovative designs are now being considered to enhance the carbide-fueled pins burnup performance of industrial cores. The first innovative design uses a buffer technology to induce a delay in getting to the fuel clad mechanical interaction. The second innovative design is a core using high plutonium content so as to optimize the fluence over burnup ratio. Preliminary results show that a burnup higher than 100 MWd/kg can be reached.As a conclusion, this global approach has proven to be efficient in revealing the benefits gained using carbide fuel in a SFR. An optimum SFR core was designed exhibiting economic competitiveness while having inherent behavior during transient and reaching high burnup performance.

The objectives of this dissertation were to find a principal domain of promising and technologically feasible reactor physics characteristics for a multi-purpose, modular-sized, lead-cooled, fast neutron spectrum reactor fueled with an advanced uranium-transuranic-nitride fuel and to determine the principal limitations for the design of an autonomous long-term multi-purpose fast reactor (ALM-FR) within the principal reactor physics characteristic domain. The objectives were accomplished by producing a conceptual design for an ALM-FR and by analysis of the potential ALM-FR performance characteristics. The ALM-FR design developed in this dissertation is based on the concept of a secure transportable autonomous reactor for hydrogen production (STAR-H2) and represents further refinement of the STAR-H2 concept towards an economical, proliferation-resistant, sustainable, multi-purpose nuclear energy system. The development of the ALM-FR design has been performed considering this reactor within the frame of the concept of a self-consistent nuclear energy system (SCNES) that satisfies virtually all of the requirements for future nuclear energy systems: efficient energy production, safety, self-feeding, non-proliferation, and radionuclide burning. The analysis takes into consideration a wide range of reactor design aspects including selection of technologically feasible fuels and structural materials, core configuration optimization, dynamics and safety of long-term operation on one fuel loading, and nuclear material non-proliferation. Plutonium and higher actinides are considered as essential components of an advanced fuel that maintains long-term operation. Flexibility of the ALM-FR with respect to fuel compositions is demonstrated acknowledging the principal limitations of the long-term burning of plutonium and higher actinides. To ensure consistency and accuracy, the modeling has been performed using state-of-the-art computer codes developed at Argonne National Laboratory. As a result of the computational analysis performed in this work, the ALM-FR design provides for the possibility of continuous operation during about 40 years on one fuel loading containing mixture of depleted uranium with plutonium and higher actinides. All reactor physics characteristics of the ALM-FR are kept within technological limits ensuring safety of ultra-long autonomous operation. The results obtained provide for identification of physical features of the ALM-FR that significantly influence flexibility of the design and its applications. The special emphasis is given to existing limitations on the utilization of higher actinides as a fuel component.

In this master thesis a small modular sodium-cooled metal-fuelled pool-type fast reactor design, called SPARC - Safe and Passive with Autonomous Reactivity control, has been designed. The long term reactivity changes in the SPARC are managed by implementation of the the Autonomous Reactivity Control (ARC) system, which is the novelty of the design. The overall design is mainly based on the Integral Fast Reactor project (IFR), which experimentally demonstrated the passive safety characteristics of a metal fuelled, sodium-cooled, pool-type reactor system. Whilst mimicking the passive safety features of the IFR, the vision of the SPARC design is a battery type reactor, which can operate with minimum interference from human actors. In this thesis, two reactor examples have been developed which operate using different fuel compositions. One reactor operates on recycled nuclear waste from today's nuclear power plants, and the other reactor operates on enriched uranium. Both reactors have a thermal power of 150 MW, and are meant to operate for 30 years without refuelling. The design was developed using the ADOPT software, and was simulated in Serpent. Using Serpent, criticality analyses were carried out which show that the ARC system is able to control the long term reactivity changes of the reactors.

The objectives of this dissertation were to find a principal domain of promising and technologically feasible reactor physics characteristics for a multi-purpose, modular-sized, lead-cooled, fast neutron spectrum reactor fueled with an advanced uranium-transuranic-nitride fuel and to determine the principal limitations for the design of an autonomous long-term multi-purpose fast reactor (ALM-FR) within the principal reactor physics characteristic domain. The objectives were accomplished by producing a conceptual design for an ALM-FR and by analysis of the potential ALM-FR performance characteristics. The ALM-FR design developed in this dissertation is based on the concept of a secure transportable autonomous reactor for hydrogen production (STAR-H2) and represents further refinement of the STAR-H2 concept towards an economical, proliferation-resistant, sustainable, multi-purpose nuclear energy system. The development of the ALM-FR design has been performed considering this reactor within the frame of the concept of a self-consistent nuclear energy system (SCNES) that satisfies virtually all of the requirements for future nuclear energy systems: efficient energy production, safety, self-feeding, non-proliferation, and radionuclide burning. The analysis takes into consideration a wide range of reactor design aspects including selection of technologically feasible fuels and structural materials, core configuration optimization, dynamics and safety of long-term operation on one fuel loading, and nuclear material non-proliferation. Plutonium and higher actinides are considered as essential components of an advanced fuel that maintains long-term operation. Flexibility of the ALM-FR with respect to fuel compositions is demonstrated acknowledging the principal limitations of the long-term burning of plutonium and higher actinides. To ensure consistency and accuracy, the modeling has been performed using state-of-the-art computer codes developed at Argonne National Laboratory. As a result of the computational analysis performed in this work, the ALM-FR design provides for the possibility of continuous operation during about 40 years on one fuel loading containing mixture of depleted uranium with plutonium and higher actinides. All reactor physics characteristics of the ALM-FR are kept within technological limits ensuring safety of ultra-long autonomous operation. The results obtained provide for identification of physical features of the ALM-FR that significantly influence flexibility of the design and its applications. The special emphasis is given to existing limitations on the utilization of higher actinides as a fuel component.

International Atomic Energy Agency (IAEA) safeguards are one method of proliferation resistance which is applied at most nuclear facilities worldwide. IAEA safeguards act to prevent the diversion of nuclear materials from a facility through the deterrence of detection. However, even with IAEA safeguards present at a facility, the country where the facility is located may still attempt to proliferate nuclear material by exploiting weaknesses in the safeguards system. The IAEA's mission is to detect the diversion of nuclear materials as soon as possible and ideally before it can be weaponized. Modern IAEA safeguards utilize unattended monitoring systems (UMS) to perform nuclear material accountancy and maintain the continuity of knowledge with regards to the position of nuclear material at a facility. This research focuses on evaluating the reliability of unattended monitoring systems and integrating the probabilistic failure of these systems into the comprehensive probabilistic proliferation resistance model of a facility. To accomplish this, this research applies reliability engineering analysis methods to probabilistic proliferation resistance modeling. This approach is demonstrated through the analysis of a safeguards design for the Example Sodium Fast Reactor Fuel Cycle Facility (ESFR FCF). The ESFR FCF UMS were analyzed to demonstrate the analysis and design processes that an analyst or designer would go through when evaluating/designing the proliferation resistance component of a safeguards system. When comparing the mean time to failure (MTTF) for the system without redundancies versus one with redundancies, it is apparent that redundancies are necessary to achieve a design without routine failures. A reliability engineering approach to probabilistic safeguards system analysis and design can be used to reach meaningful conclusions regarding the proliferation resistance of a UMS. The methods developed in this research provide analysts and designers alike a process to follow to evaluate the reliability of a UMS.

Les critères de sureté qui régissent le développement de coeurs de réacteurs dequatrième génération implique l’usage d’outils de calcul neutronique performants. Unepremière partie de la thèse reprend toutes les étapes de modélisation neutronique desréacteurs rapides actuellement d’usage dans le code de référence ECCO. La capacité desmodèles à décrire le phénomène d’autoprotection, à représenter les fuites neutroniques auniveau d’un réseau d’assemblages combustibles et à générer des sections macroscopiquesreprésentatives est appréciée sur le domaine des réacteurs rapides innovants respectant lescritères de quatrième génération. La deuxième partie de ce mémoire se consacre à lamodélisation des coeurs rapides avec réflecteur acier. Ces derniers nécessitent ledéveloppement de méthodes avancées de condensation et d’homogénéisation. Plusieursméthodes sont proposées et confrontées sur un problème de modélisation typique : le coeurZONA2B du réacteur maquette MASURCA. / Due to safety rules running on fourth generation reactors’ core development,neutronics simulation tools have to be as accurate as never before. First part of this reportenumerates every step of fast reactor’s neutronics simulation implemented in currentreference code: ECCO. Considering the field of fast reactors that meet criteria of fourthgeneration, ability of models to describe self-shielding phenomenon, to simulate neutronsleakage in a lattice of fuel assemblies and to produce representative macroscopic sections isevaluated. The second part of this thesis is dedicated to the simulation of fast reactors’ corewith steel reflector. These require the development of advanced methods of condensationand homogenization. Several methods are proposed and compared on a typical case: theZONA2B core of MASURCA reactor.