Traditionally coupled field reactor analysis has been carried out using several loosely coupled solvers, each having been developed independently from the others. In the field of multi-physics, the current generation of object-oriented toolkits provides robust close coupling of multiple fields on a single framework. This research investigates the suitability of such frameworks, in particular the Open-source Field Operation and Manipulation (OpenFOAM) framework, for the solution of spatial reactor dynamics problems. For this a subset of the theory of the Time-dependent Neutronics and Temperatures (TINTE) code, a time-dependent two-group diffusion solver, was implemented in the OpenFOAM framework. This newly created code, called diffusionFOAM, was tested for a number of steady-state and transient cases. The solver was found to perform satisfactorily, despite a number of numerical issues. The object-oriented structure of the framework allowed for rapid and efficient development of the solver. Further investigations suggest that more advanced transport methods and higher order spatial discretization schemes can potentially be implemented using such a framework as well. / Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2008.

A lattice physics code is a vital tool, forming a base of reactor coreanalysis. It enables the neutronic properties of the fuel assembly to becalculated and generates a proper set of data to be used by a 3-D full coresimulator. Due to advancement and complexity of modern Boiling WaterReactor assembly designs, a new deterministic lattice physics codeis being developed at Westinghouse Sweden AB, namely PHOENIX5.Each time a new code is written, its methodology of solving the neutrontransport equation, has to be validated to make sure it providesreliable output. In a wake of preparation for PHOENIX5 release andconsecutive validation efforts, a set of reference Monte Carlo calculationswas prepared, using the code Serpent. A depletion calculation with achosen type of branch cases was conducted. Methods implemented inPHOENIX5 are based on the Current Coupling Collision Probabilitymethod used in older versions of the code HELIOS. Therefore, a comparisonbetween reference Monte Carlo simulations and HELIOS 1.8.1is made, in order to discover problems inherent to the said method ofsolving the neutron transport equation. A special care should be givenduring PHOENIX5 validation, to issues highlighted in this work.Discrepancies in results of Serpent and HELIOS are attributed mostlyto disparities in the basic nuclear data used by the codes, as well as arange of approximations and corrections adopted by the deterministiccode.Serpent and HELIOS showed a good agreement in a typical voidrange (up to 90 % void) and ‘less’ challenging branches (coolant void,fuel temperature and spacer grid branches). More significant discrepanciesappeared for extreme cases with a very high void and control rodpresence (k1 differences as high as 1000 pcm) and rather pronouncedconcentrations of the natural boron dissolved in coolant (absolute differencesroughly at a level of 900 pcm). The issues do not seem to stemsolely from discrepancies in the nuclear data libraries used by Serpentand HELIOS.Moreover, a coolant void bias was consistently found in the resultsof branch calculation at changing coolant void. This confirms the analogousphenomenon found in previous studies of the CCCP based deterministiccodes. It most probably stems from the assumptions used bythe method while tackling the neutron transport equation, such as theflat source approximation, the isotropic scattering assumption and thetransport correction. An alternative transport correction approximationis proposed to alleviate this issue.

京都大学 / 0048 / 新制・課程博士 / 博士(エネルギー科学) / 甲第18388号 / エネ博第300号 / 新制||エネ||62(附属図書館) / 31246 / 京都大学大学院エネルギー科学研究科エネルギー変換科学専攻 / (主査)教授 小西 哲之, 教授 宇根﨑 博信, 准教授 笠田 竜太 / 学位規則第4条第1項該当 / Doctor of Energy Science / Kyoto University / DFAM

The Pressure Tube Super Critical Water Reactor (PT-SCWR) is a design with a light water coolant operating at 25 MPa above the thermodynamic critical pressure, with a separated low pressure and temperature moderator, facilitated by a High E ciency Channel consisting of a pressure tube and a porous ceramic insulator tube. The 2011 AECL reference design is considered along with a 2012 benchmark. In the 2011 reference design the coolant is permitted to ow through the insulator. The insulator region has a temperature gradient from 881 K at the inner liner tube to 478 K at the pressure tube wall. The density of light water varies by an order of magnitude depending on the local enthalpy of the uid. The lateral coolant density is estimated as a radial function at ve axial positions with the lattice physics codes WIMS-AECL and Serpent. The lateral coolant density variations in the insulator region of the PT-SCWR cause strong reactivity and CVR e ects which vary heavily on axial location due to the changes in the estimated mass of coolant and the physical relocation of the coolant closer to the moderator, as the coolant is estimated to be least dense closer to the fuel region of the coolant ow. The beta version of Serpent 2 is used to explore the lateral coolant densities in the subchannel region of the insulator in the 2012 version of the PT-SCWR. A more advanced coolant density analysis with FLUENT is used to estimate the subchannel coolant density variation, which is linked to SERPENT 2s multi-physics interface, allowing the lattice code to measure the sensitivity of the model to the analysis of the subchannels. This analysis increases the reactivity of the PT-SCWR through the displacement of the coolant. Serpent 2 is accepted as a valid lattice code for PT-SCWR analysis. / Thesis / Master of Applied Science (MASc)

Knowledge of the effects of nuclear data uncertainties and physics approximations is crucial for the development, design, operation, and accident mitigation, of nuclear power plants. A framework to create a simulated fuel bundle, based on sensitivities and similarities, has been developed. The methodology allows safe-to-handle fuel to be manufactured such that it mimics irradiated fuel and can be used to reduce simulation uncertainties and better predict an application’s response. In this work, similarity values of ck = 0.967, E = 0.992, and G = 0.891 were found between between the irradiated fuel, and non-irradiated simulated fuel. In addition, a set of ZED-2 experiments has been analyzed that are applicable to an SCWR nuclear data adjustment and simulation bias determination. This was shown through high sensitivity coverage of many important nuclides, however, a low completeness value of R=0.24 indicates the set of 39 experiments alone is not sufficient for an accurate bias determination. Lastly, a technique has been presented that reduces diffusion calculation errors through the use of novel and practical mean discontinuity factors. The discontinuity factors have shown to reduce maximum channel power errors by up to 6.7%, and reactivity errors by 2.6 mk, compared to conventional analysis techniques. / Thesis / Doctor of Philosophy (PhD) / Use of practical discontinuity factors has shown to reduce channel power predictions significantly. Furthermore, an experimental and numerical technique has been developed to improve neutron transport predictions. Finally, a set of experiments have been modeled and simulated to determine their applicability to the SCWR.

Les réacteurs de 4ème génération ont pour objectif l’avènement d’un nucléaire durable susceptible de soutenir la transition énergétique. Anticipant un possible retard, dû à des difficultés techniques et économiques, des solutions innovantes inspirées des technologies actuelles (REP) sont à l’étude. Ces réacteurs à haute conversion nécessitent le développement d’outils académiques simples et robustes adaptés aux phases de la conception et capables :- D’évaluer les performances du combustible (burnup). Cet aspect est géré par l’environnement C++ SMURE (Serpent/MCNP Utility for Reactor Evolution), ici adapté et complété pour modéliser l’évolution du combustible à l’échelle de l’assemblage comme à l’échelle du cœur.- D’évaluer les performances de sûreté, ce qui nécessite le couplage entre la neutronique, ici approximée par la théorie de la diffusion et résolue par la NDM (Nodal Drift Method), et la thermohydraulique dont le traitement est simplifié dans le code BATH (Basic Approach to ThermalHydraulics). Le couplage NDM/BATH a fait l’objet d’une validation sur un benchmark d’éjection de grappe.Nos outils et méthodes de conception sont appliqués au dimensionnement de SMR sous-modérés à eau légère fonctionnant soit au Th/U soit au Th/Pu. Outre les critères usuels de conception (i.e. facteur de forme), il s’est avéré nécessaire, pour la crédibilité du concept, de spécifier la gestion de la réactivité, ce qui a mené au développement d’une méthodologie d’optimisation des poisons consommables. L’analyse de sûreté a permis de poser de nouveaux critères de conception, notamment sur le niveau maximal de sous-modération permettant d’éviter la crise d’ébullition nucléée. Cela a également mis en lumière les implications sur la sûreté de certains choix de conception comme le fonctionnement avec un inventaire réduit en bore soluble. / The Generation IV of nuclear reactors aims at making the nuclear energy a sustainable power source, able to contribute efficiently to the energetic transition. To anticipate the delay of this Gen. IV, innovative retro-fitted nuclear reactors with high level of conversion are studied. The conception of such reactors needs the development of a flexible and robust academical tool box in order to:- Evaluate fuel performance. This is done by means of SMURE (Serpent/MCNP Utility for Reactor Evolution), the dedicated CNRS C++ framework, which is adapted to perform burnup calculation both at assembly scale and at core scale.- Evaluate safety performance. This implies coupled transient simulation between neutronics and thermohydraulics. Neutronics is handled by the Nodal Drift Method (NDM) which solves the diffusion equations while thermohydraulics is simplified and computed by the code Basic Approach to ThermalHydraulics (BATH). This coupling between NDM/BATH has been validated on a Rod Ejection Accident (REA) benchmark.Ours tools and methods are applied on the design of sub-moderated water-cooled SMR cores using either Th/U or Th/Pu fuel. In addition to basic conception criteria such as the form factor, the reactivity management has been investigated in details, which has led to the development of a new methodology for optimal used of burnable poisons. The safety analysis against REA highlights new conceptions limits, in particular on the maximal sub-moderating ratio in order to avoid nucleate boiling. It also reveals the consequences on the reactor safety of some design choices such as low soluble boron inventory.

Es wurde ein Modell zur realistischen Beschreibung der Kühlmittelvermischung innerhalb des Reaktordruckbehälters von Druckwasserreaktoren in den gekoppelten Programmkomplex DYN3D/ATHLET implementiert. Diese Modell basiert auf dem Prinzip der linearen Superposition Dirac-Impuls-ähnlicher Störungen und kann für die Modellierung der Vermischung von Kühlmittel unterschiedlicher Temperatur und/oder unterschiedlicher Borsäurekonzentration eingesetzt werden. Der um das Vermischungsmodell erweiterte Programmkomplex DYN3D/ATHLET wurde für Analyse von Borverdünnungsstörfällen und Frischdampflecks angewandt. Für den Fall "Start der ersten Hauptkühlmittelpumpe bei Vorhandensein eines minderborierten Pfropfens im kalten Strang" zeigten die Ergebnisse der durchgeführten Parameterstudie, dass es selbst bei Annahme des maximal möglichen Pfropfenvolumens nicht zu einer Schädigung des Brennstoffes kommt. Mit den Analysen zu einem generischen Frischdampfleckszenario wurde die Anwendbarkeit des Programmkomplexes DYN3D/ATHLET auf die zweite Störfallklasse, in der die Kühlmittelvermischung eine wichtige Rollen spielt, demonstriert. Im Rahmen der Arbeiten zum Projekt wurde außerdem gezeigt, dass der Einfluss der turbulenten Schwankungen des Geschwindigkeitsfeldes innerhalb des Reaktordruckbehälters auf neutronenkinetische Parameter im Nominalbetrieb und unter Störfallbedingungen nicht zu vernachlässigen ist. A model for the realistic description of the coolant mixing inside the pressure vessel of pressurized water reactors was implemented into the coupled code complex DYN3D/ATHLET. This model is based on the linear superposition Dirac-pulse-like perturbations. The model can be applied to the mixing of coolant of different temperature and/or boron concentration. The coupled code complex DYN3D/ATHLET with the newly implemented model was applied to the analysis of boron dilution and steam line break accidents. The results of a parameter study for the case "Start-up of the first main coolant pump with a slug of lower borated water in the cold leg" have shown, that even under the conditions of the maximum slug volume there is no fuel damage. The applicability to the second class of accidents, where the coolant mixing has to be considered, was demonstrated by the analysis of a generic main steam line break scenario. Further it was shown, that the influence of turbulent fluctuations of the velocity inside the reactor pressure vessel during nominal and accident conditions on neutron-kinetic parameters cannot be neglected.

Les travaux réalisés au cours de cette thèse portent sur la conception neutronique de configurations expérimentale destinées à améliorer la connaissance de paramètres neutroniques, en particulier les données nucléaires dans une gamme de spectres neutroniques élargie, dans le futur réacteur de puissance nulle ZEPHYR. La grande flexibilité expérimentale attendue de ce réacteur présente un défi majeur lié à la présence obligatoire d’une zone en eau périphérique. Or, répondre aux besoins des réacteurs électronucléaires nécessite de reproduire fidèlement leurs caractéristiques neutroniques dans les zones de mesure. L’enjeu est alors de comprendre puis réaliser une adaptation spectrale efficace entre la périphérie et la zone de mesure, et de l’appliquer pour des configurations dédiées aux réacteurs de Générations II, III et IV.Dans un premier temps les couplages thermique-rapides ont été étudiés, stimulés par le développement des réacteurs rapides de Génération IV. De telles configurations ayant été réalisées par le passé, le travail réalisé a visé à produire une démarche rigoureuse de conception permettant de s’assurer de l’indépendance neutronique de la zone centrale rapide vis-à-vis de la zone thermique périphérique. Une configuration optimisée a pu être définie et des déclinaisons permettent de cibler préférentiellement les sections efficaces d’absorption ou de diffusion lors de mesures d’oscillations d’échantillons dans un canal expérimental dédié, situé au centre du massif. Des premières études relatives à la sûreté de cette configuration optimisée ont également été réalisées afin de justifier sa faisabilité pratique.Dans un second temps, les principes d’adaptation spectrale sont appliqués pour cibler le domaine énergétique entre 10 eV et 10 keV. Ce domaine ne vise pas un soutien direct à un concept de réacteur mais à pallier au manque de mesures dans cette zone, notamment pour les aspects de sûreté/criticité.Enfin, l’étude de sensibilité aux données nucléaires des effets en réactivité mesurés au centre de ces configurations reste un problème ouvert auquel cette thèse apporte une contribution au travers d’une modélisation théorique. / This PhD thesis focuses on the neutronics design of experimental configurations to improve knowledge on neutronical parameters, in particular nuclear data in enlarged spectrum energy domains, in the future Zero Power Reactor ZEPHYR. Its awaited high spectral flexibility faces a major challenge due to the presence of a peripheral water zone. Answering the needs of current and future electronuclear reactor fleet requires a faithful reproduction of their neutronical characteristics in the measurement zone. Then, the issue is to understand how to design an efficient spectral adaptation between the periphery and the experimental zone, and to apply it on Generations II, III and IV dedicated configurations.As a first step, fast-thermal coupled cores were studied in order to support the development of Generation IV fast reactors. Such configurations had been previously realized, but our work intends to provide a rigorous design approach to insure the neutronical independence of the fast central zone regarding the peripheral thermal zone. An optimized configuration has been defined; some adaptations allow to target either absorption or scattering cross-sections during dedicated sample oscillation campaigns. Preliminary safety studies associated to the safe operation of this optimized configuration have also been realized in order to justify its practical feasibility.A second step used the spectral adaptation principles to target the 10 eV to 10 keV energy domain. Even though this domain is not directly related to a concept of reactor, it suffers from a lack of dedicated experiments despite several applications, in particular for criticality/safety issues.Finally, the sensitivity calculations of reactivity effects to nuclear data measured in the centers of these configurations still remains an unresolved problem to which this PhD thesis brings a contribution thanks to a theoretical modelling.

Es wurde ein Modell zur realistischen Beschreibung der Kühlmittelvermischung innerhalb des Reaktordruckbehälters von Druckwasserreaktoren in den gekoppelten Programmkomplex DYN3D/ATHLET implementiert. Diese Modell basiert auf dem Prinzip der linearen Superposition Dirac-Impuls-ähnlicher Störungen und kann für die Modellierung der Vermischung von Kühlmittel unterschiedlicher Temperatur und/oder unterschiedlicher Borsäurekonzentration eingesetzt werden. Der um das Vermischungsmodell erweiterte Programmkomplex DYN3D/ATHLET wurde für Analyse von Borverdünnungsstörfällen und Frischdampflecks angewandt. Für den Fall "Start der ersten Hauptkühlmittelpumpe bei Vorhandensein eines minderborierten Pfropfens im kalten Strang" zeigten die Ergebnisse der durchgeführten Parameterstudie, dass es selbst bei Annahme des maximal möglichen Pfropfenvolumens nicht zu einer Schädigung des Brennstoffes kommt. Mit den Analysen zu einem generischen Frischdampfleckszenario wurde die Anwendbarkeit des Programmkomplexes DYN3D/ATHLET auf die zweite Störfallklasse, in der die Kühlmittelvermischung eine wichtige Rollen spielt, demonstriert. Im Rahmen der Arbeiten zum Projekt wurde außerdem gezeigt, dass der Einfluss der turbulenten Schwankungen des Geschwindigkeitsfeldes innerhalb des Reaktordruckbehälters auf neutronenkinetische Parameter im Nominalbetrieb und unter Störfallbedingungen nicht zu vernachlässigen ist. A model for the realistic description of the coolant mixing inside the pressure vessel of pressurized water reactors was implemented into the coupled code complex DYN3D/ATHLET. This model is based on the linear superposition Dirac-pulse-like perturbations. The model can be applied to the mixing of coolant of different temperature and/or boron concentration. The coupled code complex DYN3D/ATHLET with the newly implemented model was applied to the analysis of boron dilution and steam line break accidents. The results of a parameter study for the case "Start-up of the first main coolant pump with a slug of lower borated water in the cold leg" have shown, that even under the conditions of the maximum slug volume there is no fuel damage. The applicability to the second class of accidents, where the coolant mixing has to be considered, was demonstrated by the analysis of a generic main steam line break scenario. Further it was shown, that the influence of turbulent fluctuations of the velocity inside the reactor pressure vessel during nominal and accident conditions on neutron-kinetic parameters cannot be neglected.

L’analyse de sûreté des réacteurs nucléaires nécessite la modélisation fine des phénomènes y survenant et plus spécifiquement ceux permettant d’assurer l’intégrité des barrières de confinement. Les outils de modélisation et codes actuels favorisent une analyse fine du système réacteur par discipline dédiée, et couplée avec des modèles simplifiés. Néanmoins, le développement depuis plusieurs années d’une approche dite « Best Estimate », basée sur des calculs multiphysiques et multi-échelle, est en cours de réalisation. Cette approche permettra d’accéder au suivi et à l’analyse détaillée de problèmes complexes tels que l’étude des Réacteurs nucléaires en situation standard et accidentelle. Dans cette approche, les phénomènes physiques sont simulés aussi précisément que possible (selon la connaissance actuelle) par les modèles couplés. Par exemple, des codes disciplinaires existent et permettent la modélisation précise de la neutronique, de la thermohydraulique du cœur du réacteur ou de la thermohydraulique sur l'ensemble du système, de la thermomécanique du combustible ou des structures. Une approche « Best Estimate » consiste à coupler ces modèles afin de réaliser une modélisation globale et précise du système de réacteur nucléaire. Cette approche nécessite de bien définir les modèles qui sont utilisés afin de préciser exactement leurs limites, et donc préciser les incertitudes des résultats des modèles couplés afin de les assumer et de les optimiser.C’est dans ce contexte de travail que s’inscrit cette thèse. Elle consiste dans le développement d'un couplage multiphysique et multi-échelle « Best Estimate » afin d'obtenir une analyse précise des Réacteurs à Eau Légère en situations normale et accidentelle. Elle a consisté principalement en l’analyse des modèles et de leurs interactions et à la mise en œuvre d'un algorithme de couplage multiphysique entre une neutronique et une thermohydraulique exprimées à l'échelle du réacteur, ainsi qu’avec une thermomécanique fine à l'échelle élémentaire du crayon combustible. En outre, un travail spécifique a été effectué afin de préparer ou d'améliorer l’accés à l'information physique locale nécessaire à la mise en œuvre de modélisations couplées multi-échelles, à l'échelle du combustible. / The safety analysis of nuclear power plants requires a deep understanding of underlying key physical phenomena that determine the integrity of the physical containment barriers. At the present time, cutting edge models focus on a single aspect (discipline) of the physical system coupled with rough models of the other aspects needed to simulate the global system. But, safety analyses can be carried out based on Multiphysics and Multiscales modelling. This Best Effort approach would give a full and accurate (High Fidelity) comprehension of the reactor core under standard and accidental situations. In this approach, the physical phenomena are simulated as accurately as possible (according to present knowledge) by coupled models in the most efficient way. For example, codes exists that are accurate modellings of Neutronics, or modellings of thermal fluid mechanics inside the core, or modellings of thermal fluid mechanics over the whole system, or modellings of thermal mechanics of the fuel pin or over the whole device structure. A Best Estimate approach would couple these models in order to realize a global and accurate modelling of the Nuclear reactor. This approach requires to define well the models that are used in order to exactly specify their limits, and hence, specify uncertainties of the coupled model results in order to assume and optimize them.It is in this context that this PhD thesis work is being under taken. It consists in the development of a Multi-physics and multi-scale Best Estimate modelling in order to obtain an accurate analysis of Pressurized Water Reactor under standard and accidental operating situations. It mainly involves the understanding of each model and their interactions, followed by the implementation of multiphysics algorithms coupling Neutronics and Thermohydraulics at reactor scale to an accurate Thermomechanics at the elementary scale of the fuel pin. In addition, a work project has been carried out in order to prepare or improve the access to the local physical informations that are needed for the implementation of multiscale coupling scheme, at the elementary scale of the fuel pin.