In this thesis, the feasibility to use sodium cooled fast reactors loaded with MOX, metallic and nitride fuels for efficient transmutation of americium is investigated by performing transient analysis for cases with different americium contents in fuels, using safety parameters obtained with the SERPENT Monte Carlo code. It was then demonstrated that there is no solid limit for the Am introduction into oxide, metallic and nitride fuels that were loaded into sodium fast reactors. Instead, higher Am contents could be permitted if specific levels of power penalty were accepted. Transient analysis of a new Accelerator Driven System design with higher neutron source efficiency than the reference EFIT-400 design, was also performed. Based on simulation results, the suggested ADS design was proved to survive the full set of transients, preserving 130 K margin to cladding rupture during the most limiting transient. After comparing Am transmutation performances in SFRs and the suggested ADS, it can be concluded that: 1. Nitride fuel could provide the highest Am transmutation efficiency, when loaded into SFRs; 2. One SFR loaded with nitride fuel is sufficient to transmute Am inventory produced by more than 15 commercial LWRs within the same time period, which is three times higher than the supporting ratio reported for the suggested ADS; 3. The total fraction of ADS power in the power park is half of cases for critical reactors. / QC 20120201

No description available.

Thesis (M. S.)--Mechanical Engineering, Georgia Institute of Technology, 2009. / Committee Chair: van Rooijen, Wilfred; Committee Member: Hertel, Nolan; Committee Member: Stacey, Weston

Der Reaktordynamikcode DYN3D wird für Kernanalysen von Natrium-gekühlten schnellen Reaktoren (SFR) erweitert. In diesem Bericht werden neu implementierte thermomechanische Modelle für die adäquate Simulation von SFR-Transienten beschrieben, die die Simulation der axialen Wärmeausdehnung von Brennstäben und die radiale Ausdehnung des Reaktorkerns umfassen. Darüber hinaus wurde das Verfahren zur Erstellung von Querschnittsbibliotheken für DYN3D für SFR-Analysen erweitert. Die Verifizierung der neuen Modelle und der Querschnittserstellung erfolgte auf Vollkern-Ebene mit stationären Experimenten von der BFS-Testanlage des IPPE Obninsk und Daten des großen oxidischen Kerns des OECD/NEA-Benchmark und den Experimenten zum Zyklusende des Phenix-Kerns. Die DYN3D-Ergebnisse wurden mit der Monte-Carlo-Referenzlösung verglichen, die durch den SERPENT-Code berechnet wurde. Die Testergebnisse zeigen, dass die neu entwickelten Modelle die Wärmeausdeh-nungseffekte der Kernstruktur genau berücksichtigen können. Das neu entwickelte Verfahren zur Erstellung von Querschnittsbibliotheken wurde ebenfalls auf der Basis von SERPENT-Ergebnissen erfolgreich verifiziert. Zur Validierung wurden mehrere Tests, die sowohl stationäre als auch transiente Fälle aus den Phenix-Experimenten enthalten, mit DYN3D berechnet. Die DYN3D-Lösungen weisen eine gute Übereinstimmung mit den experimentellen Daten auf, was die Anwendbarkeit der Codes für Kernanalysen von Natrium-gekühlten schnellen Reaktoren bestätigt.

The sensitivity of two operational output parameters, criticality and isotopic composition during burnup, to specific design and operational reactor parameters in a Sodium Cooled Fast Reactor, is investigated. The computational simulation tool Serpent is used. The parameters varied include Uranium enrichment, Plutonium content, rod thickness, fuel temperature, and sodium density. In burnup, the development of the fraction of fissile isotopes, isotopes used for measurements, the isotopic composition of Plutonium, and isotopes that complicate fuel reprocessing is displayed. A surrogate model, optimized for use in determining how criticality develops between data points, is used. The results are displayed as plots created in Matlab. The results are discussed, with a focus on how large an effect varying different parameters have on different outputs related to the reactor's operation. It is concluded that the Plutonium content has the largest effect on the isotopic composition and that, based on the performed simulations, MOX fuel is potentially safer than Zirconium alloy fuel in a practical setting.

Der Reaktordynamikcode DYN3D wird für Kernanalysen von Natrium-gekühlten schnellen Reaktoren (SFR) erweitert. In diesem Bericht werden neu implementierte thermomechanische Modelle für die adäquate Simulation von SFR-Transienten beschrieben, die die Simulation der axialen Wärmeausdehnung von Brennstäben und die radiale Ausdehnung des Reaktorkerns umfassen. Darüber hinaus wurde das Verfahren zur Erstellung von Querschnittsbibliotheken für DYN3D für SFR-Analysen erweitert. Die Verifizierung der neuen Modelle und der Querschnittserstellung erfolgte auf Vollkern-Ebene mit stationären Experimenten von der BFS-Testanlage des IPPE Obninsk und Daten des großen oxidischen Kerns des OECD/NEA-Benchmark und den Experimenten zum Zyklusende des Phenix-Kerns. Die DYN3D-Ergebnisse wurden mit der Monte-Carlo-Referenzlösung verglichen, die durch den SERPENT-Code berechnet wurde. Die Testergebnisse zeigen, dass die neu entwickelten Modelle die Wärmeausdeh-nungseffekte der Kernstruktur genau berücksichtigen können. Das neu entwickelte Verfahren zur Erstellung von Querschnittsbibliotheken wurde ebenfalls auf der Basis von SERPENT-Ergebnissen erfolgreich verifiziert. Zur Validierung wurden mehrere Tests, die sowohl stationäre als auch transiente Fälle aus den Phenix-Experimenten enthalten, mit DYN3D berechnet. Die DYN3D-Lösungen weisen eine gute Übereinstimmung mit den experimentellen Daten auf, was die Anwendbarkeit der Codes für Kernanalysen von Natrium-gekühlten schnellen Reaktoren bestätigt.

No description available.

This thesis is a follow up on "SPARC fast reactor design: Design of two passively metal-fuelled sodium-cooled pool-type small modular fast reactors with Autonomous Reactivity Control" by Tobias Lindström (2015). In this thesis the two reactors designed by Lindström in said thesis were evaluated. The goal was to determine the reactors ability to load follow as well as the burnup of the neutron absorber used in the passive control system. To be able to determine the dynamic behaviour of the reactors the reactivity feedbacks of the cores were modelled using Serpent, a Monte Carlo simulation software for 3D neutron transport calculations. These feedbacks were then implemented into a dynamic simulation of the core, primary and secondary circulation and steam generator. The secondary circulation and feedwater flow were used to regulate steam temperature and turbine power. The core was left at constant coolant flow and no control rods were used. The simulations showed that the reactor was able to load follow between 100 % and 40 % of rated power at a speed of 6 % per minute. It was also shown that the reactor could safely adjust its power between 100 % and 10 % of rated power suggesting that load following is possible below 40 % of rated power but at a lower speed. Finally the reactors were allowed compensate for the variations in a week of the Latvian wind power production in order to show one possible application of the reactor.

Safety and reliability are one of the key objectives for future Generation IV nuclear energy systems. The neutron flux monitoring system forms an integral part of the safety design of a nuclear reactor and must be able to detect any irregularities during all states of reactor operation. The work in this thesis mainly concerns the detection of in-core perturbations arising from unwanted movements of control rods with in-vessel neutron detectors in a sodium-cooled fast reactor. Feasibility study of self-powered neutron detectors (SPNDs) with platinum emitters as in-core power profile monitors for SFRs at full power is performed. The study shows that an SPND with a platinum emitter generates a prompt current signal induced by neutrons and gammas of the order of 600 nA/m, which is large enough to be measurable. Therefore, it is possible for the SPND to follow local power fluctuations at full power operation. Ex-core and in-core detector locations are investigated with two types of detectors, fission chambers and self-powered neutron detectors (SPNDs) respectively, to study the possibility of detection of the spatial changes in the power profile during two different transient conditions, i.e. inadvertent withdrawal of control rods (IRW) and one stuck rod during reactor shutdown (OSR). It is shown that it is possible to detect the two simulated transients with this set of ex-core and in-core detectors before any melting of the fuel takes place. The detector signal can tolerate a noise level up to 5% during an IRW and up to 1% during an OSR.

Les barres de commande du futur démonstrateur de réacteur à neutrons rapides refroidi au sodium (RNR – Na) nommé ASTRID sont constituées de pastilles de B4C enfermées dans une gaine en acier inoxydable AIM1 (15Cr-15Ni-0,4Ti). En service, les pastilles de B4C sont plongées dans le sodium liquide à une température allant de 500 à 600 °C. Les retours d’expérience des RNR - Na ont mis en évidence que la durée de vie des barres de commande était limitée par leur cinétique de carburation. Cependant, un phénomène de boruration des gaines a été observé lors d’essais réalisés « hors réacteur / hors irradiation ». Afin de maîtriser la durabilité des barres de commandes, il est donc nécessaire d’évaluer précisément la nature de l’interaction entre les gaines en acier et le B4C dans le sodium liquide. Ainsi, deux campagnes d’essai ont été menées : 1. Trois aciers inoxydables (AIM1, 316L et EM10) ont été exposés dans du sodium liquide fortement carburant (ac > 1) à 500, 600 et 650 °C. 2. Les mêmes nuances d’aciers ont été exposées dans du sodium liquide contenant de la poudre de B4C en excès à 500 et 600 °C. La première campagne a été réalisée pour avoir une meilleure compréhension des mécanismes et des cinétiques de carburation des barres de commande. Tout d’abord, l’état de carburation a été caractérisé finement au moyen de différentes techniques d’analyse (microsonde de Castaing, diffraction des rayons X du rayonnement synchrotron, microscopie électronique en transmission). Ensuite, la cinétique de carburation a été simulée à l’aide d’un modèle analytique simplifié de la carburation puis grâce à un outil commercial plus complet de simulation numérique de la diffusion à l’équilibre thermodynamique (DICTRA). Des écarts ont été observés entre les simulations des états de carburation réalisées avec DICTRA et les mesures expérimentales (profil de concentration en carbone et population de carbures). Afin de prédire au mieux l’état de carburation des aciers rencontré à 500 et 600 °C, il a notamment été démontré qu’il est probablement nécessaire de prendre en compte la diffusion du carbone dans les joints de grains et un écart à l’équilibre thermodynamique entre le carbone piégé dans les carbures et le carbone dissout dans la matrice. La deuxième campagne expérimentale a concerné l’étude du système : acier – B4C – Na. Des caractérisations couplées à des études thermodynamique et cinétique ont permis de proposer un mécanisme de carburation et de boruration des aciers. Après la dissolution du B4C dans le sodium, deux phénomènes ont été observés. Le bore réagit avec les aciers pour former une couche duplexe de borures à la surface (MB, M2B) et des borures dans les joints de grains du substrat. La cinétique de formation de la couche de borures dans les aciers suit une loi parabolique. Le carbone entraine une légère carburation des aciers plus en profondeur et le degré de carburation des aciers s’est avéré constant entre 250 et 3000 h d’exposition, ce qui suggère que le phénomène de carburation s’opère probablement avant la formation d’une couche continue de borures. / Pellets of boron carbide, B4C, enclosed in AIM1 (15Cr-15Ni-0.4Ti) stainless steel tubes are constitutive materials of the control rods in the future French Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID). During reactor operation, the B4C pellets are immersed in liquid sodium in the temperature range 773-873 K. Based on the feedback from operation of former Sodium Fast Reactors (SFR), the lifetime of the control rods has been shown to be limited by their carburization kinetics. Although, boriding of the steels was observed in out-ofpile studies. In order to increase the lifetime prediction of the aforementioned components in service, detailed information on the chemical interaction between the steel and B4C in liquid sodium is required. As a result, two sets of out-of-pile experiments were conducted: 1. Three stainless steels (AIM1, 316L, EM10) were exposed to highly carburizing sodium (ac > 1) at 773, 873 and 923 K. 2. The same grades were exposed to high purity B4C powder in liquid sodium at 773 and 873 K. The first campaign was performed in order to have a better understanding of the carburization phenomenology and kinetics of the control rods. The extent of carburization was evaluated. A good description of the carburization kinetics was obtained by means of two models and a simulation tool (DICTRA). The limits of the simulation tools were exposed. It was shown that the grain boundary diffusion of carbon had to be taken into account. The second set of experiments was carried out in order to study the system: steel – B4C – Na. A thorough examination of the nature of the chemical interaction was performed. The characterizations were combined with a thermodynamic and kinetic study to propose a carburization and boriding mechanism. The B4C powder dissolved in liquid sodium and reacted with the steels to form a boride layer (MB and M2B) at the surface, borides in the grain boundaries and a carburized zone underneath. The growth kinetics of the boron affected zone was shown to be parabolic. The carburization depth did not evolve between 250 and 3000 h and suggested that this phenomenon occurred during a transient stage.