Ziel dieser Arbeit war es, die Auswirkungen der im Kühlmittel von Leichtwasserreaktoren zur Reaktivitätssteuerung eingesetzten Borverbindungen auf Siedeprozesse – und damit indirekt auf die Wärmeabfuhr der Brennelemente – zu untersuchen. Bei den Siedeversuchen, die Gegenstand der vorliegenden Arbeit sind, wurde besonders auf eine realitätsnahe Annäherung an die Reaktorparameter Wert gelegt. Als Unterstützung zur Interpretation der Ergebnisse dienten eigene Messungen von signifikanten physikalischen Stoffdaten an wässrigen Borsäure- und Pentaboratlösungen. Die Siedeprozesse wurden in einer eigens für diese Analysen konzipierten und errichteten Versuchsanlage SECA unter Verwendung eines Leitfähigkeitsgittersensors sowie einer Hochgeschwindigkeitskamera bei Drücken von maximal 40 bar und Temperaturen bis zu 250 °C untersucht. Entsprechend der in den Untersuchungen gewonnenen Erkenntnis wird für reale Reaktoren fol-gendes angenommen: Die Anwesenheit von Borsäure hat keinen Einfluss auf großvolumige Sie-devorgänge im betrachteten Störfallszenario eines Druckwasserreaktors, und die Auswirkungen auf das unterkühlte Sieden sind vernachlässigbar gering. Es ist nicht zu erwarten, dass der Wärmeübergang von den Brennelementen an das Kühlmittel beeinflusst wird. Bei einer Einspeisung von Pentaborat in Siedewasserreaktoren kann jedoch davon ausgegangen werden, dass der Wärmeübergang durch eine Verkleinerung der Blasen verbessert wird. Weitere Untersuchungen bezüglich des Austrages von Pentaborat an der Phasengrenze sowie der Bildung von Schäumen sind jedoch notwendig, und es ist den Fragen nachzugehen, ob sich diese Schäume auch bei der Einspeisung von Pentaborat in einen Siedewasserreaktor bilden können und welche Auswirkungen diese auf die oberhalb des Kerns befindlichen Dampfabscheiderzyklone und Dampftrockner haben.

This master's thesis is an investigation and evaluation of MELCOR (a software tool for severe accident analyses regarding nuclear power plants), or more correctly of the (ASEA-Atom BWR 75) reactor model developed for version 1.8.6 of MELCOR. The main objective was to determine if MELCOR, with the reactor model in question, is able to produce satisfactory results in severe accident analyses compared to results made by MAAP, which is currently the only official software tool for this application in Sweden. The thesis work is related to the RASTEP project. This project has been carried out in several stages on behalf of SSM since 2009, with a number of specific issues explored within an NKS funded R&D project carried out 2011-2013. This investigation is related to the NKS part of the project. The purpose with the RASTEP project is to develop a method for rapid source term prediction that could aid the authorities in decision making during a severe accident in a nuclear power plant. A software tool, which also gave the project its name, i.e. RASTEP (RApid Source TErm Prediction), is therefore currently under development at Lloyd's Register Consulting. A software tool for severe accident analyses is needed to calculate the source terms which are the end result from the predictions made by RASTEP. A set of issues have been outlined in an earlier comparison between MAAP and MELCOR. The first objective was therefore to resolve these pre-discovered issues, but also to address new issues, should they occur. The existing MELCOR reactor model also had to be further developed through the inclusion of various safety systems, since these systems are required for certain types of scenarios. Subsequently, a set of scenarios was simulated to draw conclusions from the additions made to the reactor model. Most of the issues (pre-discovered as well as new ones) could be resolved. However the work also rendered a set of issues which are in need of further attention and investigation. The overall conclusion is that MELCOR is indeed a promising alternative for severe accident analyses in the Swedish work with nuclear safety. Several potential benefits from making use of MELCOR besides MAAP have been identified. In conclusion, they would be valuable assets to each other, e.g. since deviations in the results (between the two codes) would highlight possible weaknesses of the simulations. Finally it is recommended that the work on improving the MELCOR reactor model should continue. / RASTEP

As the United States expands its quantity of nuclear reactors in the near future, the amount of spent nuclear fuel (SNF) will also increase. Closing the nuclear fuel cycle has become the next major technical challenge for the nuclear energy industry. By separating the transuranics (TRU) from the SNF discharged by Light Water Reactors, it is possible to fuel Advanced Burner Reactors to minimize the amount of SNF that must be stored in High Level Waste Repositories. One such ABR concept is the Subcritical Advanced Burner Reactor (SABR) being developed at the Georgia Institute of Technology. SABR is a subcritical, sodium-cooled fast reactor with a fusion neutron source capable of burning up to 25% of the TRU fuel over an 8.2 year residence time. In the SABR concept an annular core with a thickness of 0.6 m and an active height of 3.2 m surrounds the toroidal fusion neutron source. Neutron multiplication varies during the lifetime of the reactor from keff = 0.95 at the beginning of reactor life to 0.83 at the end of an equilibrium fuel cycle. Sixteen control rods worth 9$ are symmetrically positioned around the reactor. This thesis describes the dynamic safety analysis of the coupled neutron source, reactor core and reactor heat removal systems. A special purpose simulation model was written to predict steady-state conditions and accident scenarios in SABR by calculating the coupled evolution of the power output from the fusion and fission cores and the axial and radial temperature distributions of a fuel pin in the reactor. Reactivity Feedback was modeled for Doppler and sodium coolant voiding. SABR has a positive temperature reactivity feedback coefficient. A series of accident scenarios were simulated to determine how much time exists to implement corrective measures during an accident before damage to the reactor occurs.

La conception du coeur d’un réacteur nucléaire est fortement multidisciplinaire (neutronique, thermo-hydraulique, thermomécanique du combustible, physique du cycle, etc.). Le problème est aussi de type multi-objectif (plusieurs performances) à grand nombre de dimensions (plusieurs dizaines de paramètres de conception).Les codes de calculs déterministes utilisés traditionnellement pour la caractérisation des coeurs demandant d’importantes ressources informatiques, l’approche de conception classique rend difficile l’exploration et l’optimisation de nouveaux concepts innovants. Afin de pallier ces difficultés, une nouvelle méthodologie a été développée lors de ces travaux de thèse. Ces travaux sont basés sur la mise en oeuvre et la validation de schémas de calculs neutronique et thermo-hydraulique pour disposer d’un outil de caractérisation d’un coeur de réacteur à neutrons rapides à caloporteur sodium tant du point de vue des performances neutroniques que de son comportement en transitoires accidentels.La méthodologie mise en oeuvre s’appuie sur la construction de modèles de substitution (ou métamodèles) aptes à remplacer la chaîne de calcul neutronique et thermo-hydraulique. Des méthodes mathématiques avancées pour la planification d’expériences, la construction et la validation des métamodèles permettent de remplacer cette chaîne de calcul par des modèles de régression au pouvoir de prédiction élevé.La méthode est appliquée à un concept innovant de coeur à Faible coefficient de Vidange sur un très large domaine d’étude, et à son comportement lors de transitoires thermo-hydrauliques non protégés pouvant amener à des situations incidentelles, voire accidentelles. Des analyses globales de sensibilité permettent d’identifier les paramètres de conception influents sur la conception du coeur et son comportement en transitoire. Des optimisations multicritères conduisent à des nouvelles configurations dont les performances sont parfois significativement améliorées. La validation des résultats produits au cours de ces travaux de thèse démontre la pertinence de la méthode au stade de la préconception d’un coeur de réacteur à neutrons rapides refroidi au sodium. / Nuclear reactor core design is a highly multidisciplinary task where neutronics, thermal-hydraulics, fuel thermo-mechanics and fuel cycle are involved. The problem is moreover multi-objective (several performances) and highly dimensional (several tens of design parameters).As the reference deterministic calculation codes for core characterization require important computing resources, the classical design method is not well suited to investigate and optimize new innovative core concepts. To cope with these difficulties, a new methodology has been developed in this thesis. Our work is based on the development and validation of simplified neutronics and thermal-hydraulics calculation schemes allowing the full characterization of Sodium-cooled Fast Reactor core regarding both neutronics performances and behavior during thermal hydraulic dimensioning transients.The developed methodology uses surrogate models (or metamodels) able to replace the neutronics and thermal-hydraulics calculation chain. Advanced mathematical methods for the design of experiment, building and validation of metamodels allows substituting this calculation chain by regression models with high prediction capabilities.The methodology is applied on a very large design space to a challenging core called CFV (French acronym for low void effect core) with a large gain on the sodium void effect. Global sensitivity analysis leads to identify the significant design parameters on the core design and its behavior during unprotected transient which can lead to severe accidents. Multi-objective optimizations lead to alternative core configurations with significantly improved performances. Validation results demonstrate the relevance of the methodology at the predesign stage of a Sodium-cooled Fast Reactor core.

Ziel dieser Arbeit war es, die Auswirkungen der im Kühlmittel von Leichtwasserreaktoren zur Reaktivitätssteuerung eingesetzten Borverbindungen auf Siedeprozesse – und damit indirekt auf die Wärmeabfuhr der Brennelemente – zu untersuchen. Bei den Siedeversuchen, die Gegenstand der vorliegenden Arbeit sind, wurde besonders auf eine realitätsnahe Annäherung an die Reaktorparameter Wert gelegt. Als Unterstützung zur Interpretation der Ergebnisse dienten eigene Messungen von signifikanten physikalischen Stoffdaten an wässrigen Borsäure- und Pentaboratlösungen. Die Siedeprozesse wurden in einer eigens für diese Analysen konzipierten und errichteten Versuchsanlage SECA unter Verwendung eines Leitfähigkeitsgittersensors sowie einer Hochgeschwindigkeitskamera bei Drücken von maximal 40 bar und Temperaturen bis zu 250 °C untersucht. Entsprechend der in den Untersuchungen gewonnenen Erkenntnis wird für reale Reaktoren fol-gendes angenommen: Die Anwesenheit von Borsäure hat keinen Einfluss auf großvolumige Sie-devorgänge im betrachteten Störfallszenario eines Druckwasserreaktors, und die Auswirkungen auf das unterkühlte Sieden sind vernachlässigbar gering. Es ist nicht zu erwarten, dass der Wärmeübergang von den Brennelementen an das Kühlmittel beeinflusst wird. Bei einer Einspeisung von Pentaborat in Siedewasserreaktoren kann jedoch davon ausgegangen werden, dass der Wärmeübergang durch eine Verkleinerung der Blasen verbessert wird. Weitere Untersuchungen bezüglich des Austrages von Pentaborat an der Phasengrenze sowie der Bildung von Schäumen sind jedoch notwendig, und es ist den Fragen nachzugehen, ob sich diese Schäume auch bei der Einspeisung von Pentaborat in einen Siedewasserreaktor bilden können und welche Auswirkungen diese auf die oberhalb des Kerns befindlichen Dampfabscheiderzyklone und Dampftrockner haben.

DYN3D is an best estimate advanced code for the three-dimensional simulation of steady-states and transients in light water reactor cores with quadratic and hexagonal fuel assemblies. Burnup and poison-dynamic calculations can be performed. For the investigation of wide range transients, DYN3D is coupled with system codes as ATHLET and RELAP5. The neutron kinetic model is based on the solution of the three-dimensional two-group neutron diffusion equation by nodal expansion methods. The thermal-hydraulics comprises a one- or two-phase coolant flow model on the basis of four differential balance equations for mass, energy and momentum of the two-phase mixture and the mass balance for the vapour phase. Various cross section libraries are linked with DYN3D. Systematic code validation is performed by FZR and independent organizations.

No description available.

This project deals with two-phase CFD model validation, a subject which is currently under active research due to its complexity. The goal is to create a model that predicts data profiles to an acceptable degree for a wide array of flow conditions. The applications within the nuclear field would mainly be for safety analysis, e.g. to predict phenomena such as the critical heat flux. The underlying physics were investigated, and simulations were performed of two-phase flow of the coolant R12 using the program OpenFOAM in an attempt to match radial profiles of void fraction, interfacial area concentration, vapour velocity and sauter mean diameter for different flow conditions provided by the DEBORA experiments. The proper set of models was found via sensitivity testing: trying combinations of different models and model coefficients. The effect on the simulation result was investigated, with the models that improved the result kept while the rest were discarded. The main strategy was to find models that accurately predicted the sauter mean diameter, as initial sensitivity tests showed the result to be highly dependent on this parameter. The impact of initial conditions and mesh refinement was also investigated, and a temperature validation study was done. The process was aided by a number of Matlab programs written by the author, to calculate and verify inputs and to post-process the result. A model was found that simulated the subcooled nucleate boiling datasets to an acceptable degree. The model failed to accurately simulate saturated nucleate boiling. / Detta projekt behandlar tvåfas CFD-modellvalidering, ett ämne som fortfarande ärunder aktiv forskning på grund av sin komplexitet. Syftet är att utveckla en modellsom förutser dataprofiler till en acceptabel grad för en mängd olika flödesregimer.Inom kärnkraftsbranchen skulle detta primärt vara applicerbart inom reaktorsäkerhet,t.ex. för att förutse fenomen så som critical heat flux. Den underliggande fysikenundersöktes, och simuleringar genomfördes på tvåfas flöde av kylmedlet R12 med hjälpav programmet OpenFOAM i ett försök att matcha 14 dataprofiler för void fraction,interfacial area, vapour velocity och sauter mean diameter för olika flödesregimertillhandahålla av DEBORA-experimenten.Den rätta uppsättningen modeller fanns via känslighetsanalys, genom att testa olikakombinationer av modeller och modellkoefficienter. Deras effektersimuleringsresultatet undersöktes, och de modeller som förbättrade resultatet behölls,medans resten förkastades. Huvudstrategin var att hitta modeller som förutsåg sautermean diameter, bubblornas medeldiameter, väl, då preliminär känslighetsanalysvisade att resultaten var mycket känsliga på denna parameter. Inflytandet avinitialvillkor och mesh-förfining undersöktes, och en temperaturprofilsvalideringgenomfördes. Till hjälp i processern var ett antal Matlab-program skrivna avförfattaren för att beräkna och verifiera inmatning och behandla och visualiserautdatan. En modell hittades som simulerade underkyld kokning till en acceptabel grad.Modellen misslyckades med att träffsäkert simulera mättad kokning.

[ES] Ante el desafío que implica la reducción de los efectos del cambio climático, la industria nuclear se ha postulado como una buena alternativa para sustituir la producción de energía eléctrica a partir de combustibles fósiles. No obstante, debe constatar la seguridad de las centrales, para lo que resulta indispensable poder predecir su comportamiento ante escenarios operacionales y accidentales. A tal efecto, y dada la imposibilidad de disponer de datos de planta para analizar estos transitorios, se generan bases de datos en instalaciones a escala reducida a partir de experimentos, siendo necesarios métodos y estrategias de escalado que permitan extrapolar los comportamientos termohidráulicos. Pese a la relevante contribución que suponen los experimentos al campo de la seguridad nuclear, en ocasiones se cuestiona la validez de sus resultados para reproducir el comportamiento de las centrales. Este hecho motiva la ejecución de test counterpart entre distintas instalaciones, que contribuyen a abordar la problemática del escalado, así como a demostrar la adecuación de los códigos termohidráulicos para predecir una respuesta realista de los sistemas. La presente tesis doctoral explora la posibilidad de aumentar el número de experimentos counterpart a partir de la definición de nuevos escenarios y su simulación con el código termohidráulico TRACE5. Con este fin, se han desarrollado modelos de las instalaciones ATLAS y LSTF, y se han estudiado y simulado experimentos counterpart ya existentes entre dichas instalaciones. La identificación de los fenómenos termohidráulicos más significativos, y el análisis de su escalado y distorsión, configuran la base de conocimientos para abordar el diseño de los nuevos test. En la tesis, en particular, se plantea un escenario tipo station blackout para LSTF partiendo de las condiciones iniciales y de contorno de un test previo en ATLAS. La simulación del experimento confirma la idoneidad de ATLAS y LSTF para realizar experimentos counterpart, en los que la fenomenología relevante es similar, y pone de manifiesto algunas limitaciones de estas instalaciones en cuanto a la extrapolabilidad de ciertos fenómenos, debido a las distorsiones originadas por la diferencia de escala y tecnología. / [CA] Davant del desafiament que implica la reducció dels efectes del canvi climàtic, la indústria nuclear s'ha postulat com una bona alternativa per a substituir la producció d'energia elèctrica a partir de combustibles fòssils. No obstant això, ha de constatar la seguretat de les centrals, per al que resulta indispensable poder predir el seu comportament davant d'escenaris operacionals i accidentals. A aquest efecte, i donada la impossibilitat de disposar de dades de planta per a analitzar aquests transitoris, es generen bases de dades en instal·lacions a escala reduïda a partir d'experiments, sent necessaris mètodes i estratègies d'escalat que permeten extrapolar els comportaments termohidràulics. Malgrat la rellevant contribució que suposen els experiments al camp de la seguretat nuclear, de vegades es qüestiona la validesa dels seus resultats per a reproduir el comportament de les centrals. Aquest fet motiva l'execució de test counterpart entre distintes instal·lacions, que contribuïxen a abordar la problemàtica de l'escalat, així com a demostrar l'adequació dels codis termohidràulics per a predir una resposta realista dels sistemes. La present tesi doctoral explora la possibilitat d'augmentar el nombre d'experiments counterpart a partir de la definició de nous escenaris i la seua simulació amb el codi termohidràulic TRACE5. Amb aquest fi, s'han desenvolupat models de les instal·lacions ATLAS i LSTF, i s'han estudiat i simulat experiments counterpart ja existents entre les dites instal·lacions. La identificació dels fenòmens termohidràulics més significatius, i l'anàlisi del seu escalat i distorsió, configuren la base de coneixements per a abordar el disseny dels nous test. En la tesi, en particular, es planteja un escenari tipus station blackout per a LSTF partint de les condicions inicials i de contorn d'un test previ en ATLAS. La simulació de l'experiment confirma la idoneïtat d'ATLAS i LSTF per a realitzar experiments counterpart, en els que la fenomenologia rellevant és semblant, i posa de manifest algunes limitacions d'aquestes instal·lacions quant a l'extrapolabilitat de certs fenòmens, a causa de les distorsions originades per la diferència d'escala i tecnologia. / [EN] Faced with the challenge of reducing the effects of climate change, the nuclear industry has been postulated as a good alternative to replace the production of electricity from fossil fuels. However, it must verify the safety of the plants, for which it is essential to be able to predict their behavior in operational and accidental scenarios. To this end, and given the impossibility of having plant data to analyze these transients, databases are generated in reduced-scale facilities from experiments, being necessary scaling methods and strategies that allow the extrapolation of thermohydraulic behaviors. Despite the relevant contribution that experiments make to the field of nuclear safety, the validity of their results to reproduce the behavior of plants is sometimes questioned. This fact motivates the execution of counterpart tests between different facilities, which contribute to addressing scaling issues, as well as to demonstrate the adequacy of the thermal-hydraulic codes to predict a realistic response of the systems. This Ph.D. Thesis explores the possibility of increasing the number of counterpart experiments based on the definition of new scenarios and their simulation with the TRACE5 thermal-hydraulic code. In order to achieve this goal, models of the ATLAS and LSTF facilities have been developed, and counterpart experiments already existing between these facilities have been studied and simulated. The identification of the most significant thermal-hydraulic phenomena and the analysis of their scaling and distortion, configure the knowledge basis to approach the design of the new tests. In the Thesis, in particular, a station blackout scenario for LSTF based on the initial and boundary conditions of a previous test in ATLAS is proposed. The simulation of the experiment confirms the suitability of ATLAS and LSTF to perform counterpart experiments, in which the relevant phenomenology is similar. Moreover, it reveals some limitations of these facilities in terms of the extrapolability of certain phenomena, due to the distortions caused by the difference in scale and technology. / Lorduy Alós, M. (2022). Análisis termohidráulico de la instalación ATLAS. Aplicaciones de la metodología de escalado [Tesis doctoral]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/181700

Perhaps surprisingly, the largest experience in operating nuclear power plants has been in nuclear naval propulsion, particularly submarines. This accumulated experience may become the basis of a proposed new generation of compact nuclear power plant designs. In an effort to de-carbonise commercial freight shipping, there is growing interest in the possibility of using nuclear propulsion systems. Reactor cores for such an application would need to be fundamentally different from land-based power generation systems, which require regular refueling, and from reactors used in military submarines, as the fuel used could not conceivably be as highly enriched. Nuclear-powered propulsion would allow ships to operate with low fuel costs, long refueling intervals, and minimal emissions; however, currently such systems remain largely confined to military vessels. This research project undertakes computational modeling of possible soluble-boron-free (SBF) reactor core designs for this application, with a view to informing design decisions in terms of choices of fuel composition, materials, core geometry and layout. Computational modeling using appropriate reactor physics (e.g. WIMS, MONK, Serpent and PANTHER), thermal-hydraulics etc. codes (e.g. COBRA-EN) is used for this project. With an emphasis on reactor physics, this study investigates possible fuel assembly and core designs for civil marine propulsion applications. In particular, it explores the feasibility of using uranium/thorium-rich fuel in a compact, long-life reactor and seek optimal choices and designs of the fuel composition, reactivity control, assembly geometry, and core loading in order to meet the operational needs of a marine propulsion reactor. In this reactor physics and 3D coupled neutronics/thermal-hydraulics study, we attempt to design a civil marine reactor core that fulfills the objective of providing at least 15 effective full-power-years (EFPY) life at 333 MWth. In order to unleash the benefit of thorium in a long life core, the micro-heterogeneous ThO2-UO2 duplex fuel is well-positioned to be utilized in our proposed civil marine core. Unfortunately, A limited number of studies of duplex fuel are available in the public domain, but its use has never been examined in the context of a SBF environment for long-life small modular rector (SMR) core. Therefore, we assumed micro-heterogeneous ThO2-UO2 duplex fuel for our proposed marine core in order to explore its capability. For the proposed civil marine propulsion core design, this study uses 18% U-235 enriched micro-heterogeneous ThO2-UO2 duplex fuel. To provide a basis for comparison we also evaluate the performance of homogeneously mixed 15% U-235 enriched all-UO2 fuel. This research also attempts to design a high power density core with 14 EFPY while satisfying the neutronic and thermal-hydraulics safety constraints. A core with an average power density of 100 MW/m3 has been successfully designed while obtaining a core life of 14 years. The average core power density for this core is increased by ∼50% compared to the reference core design (63 MW/m3 and is equivalent to Sizewell B PWR (101.6 MW/m3 which means capital costs could be significantly reduced and the economic attractiveness of the marine core commensurately improved. In addition, similar to the standard SMR core, a reference core with a power density of 63 MW/m3 has been successfully designed while obtaining a core life of ∼16 years. One of the most important points that can be drawn from these studies is that a duplex fuel lattice needs less burnable absorber than uranium-only fuel to achieve the same poison performance. The higher initial reactivity suppression and relatively smaller reactivity swing of the duplex can make the task of reactivity control through BP design in a thorium-rich core easier. It is also apparent that control rods have greater worth in a duplex core, reducing the control material requirements and thus potentially the cost of the rods. This research also analyzed the feasibility of using thorium-based duplex fuel in different cases and environments to observe whether this fuel consistently exhibit superior performance compared to the UO2 core in both the assembly and whole-core levels. The duplex fuel/core consistently exhibits superior performance in consideration of all the neutronic and TH constraints specified. It can therefore be concluded from this study that the superior performance of the thorium-based micro-heterogeneous ThO2-UO2 duplex fuel provides enhanced confidence that this fuel can be reliably used in high power density and long-life SBF marine propulsion core systems, offering neutronic advantages compared to the all-UO2 fuel. Last, but not least, considering all these factors, duplex fuel can potentially open the avenue for low-enriched uranium (LEU) SBF cores with different configurations. Motivated by growing environmental concerns and anticipated economic pressures, the overall goal of this study is to examine the technological feasibility of expanding the use of nuclear propulsion to civilian maritime shipping and to identify and propose promising candidate core designs.