In this study, liquid-vapor vertical upflow has been research with the intent of finding an improved method of modelling the interphase friction in two-phase vertical flow in nuclear thermal-hydraulic codes. An improved method of modelling interphase friction should allow for better prediction of pressure gradient, void fraction and the phasic velocities. Data has been acquired from several available published resources and analyzed to determine the interphase friction using a force balance between the liquid and vapor phases. Using the Buckingham Pi Theorem, a dimensionless interphase friction force was tested and refined before being compared against seven other dimensionless parameters. Three correlations have been developed that establish a dimensionless interphase friction force as a function of the Weber number, the Froude number and the mixture Froude number. Statistical analysis of the three correlations shows that the mixture Froude number correlation should be the most accurate correlation. The correlations have a weakness that makes them ineffective mostly for bubbly flow and some slug flow scenarios, while they should perform significantly better for annular flow cases. Comparisons have been made against the interphase friction calculations published in the manuals of RELAP5/MOD2, RELAP5/MOD3.3, RELAP5-3D and TRACE. The findings have generally shown that the equations in the manuals provide very inaccurate approximations of the interphase friction compared to the interphase friction that was found via force balance. When analyzing the source code of RELAP5/MOD3.3, several differences were noticed between the source code and manual, which have been discussed. Calculations with the source code equations reveal that the source code provides a modestly improved prediction of the interphase friction force, but still has significant errors. Despite the fact that the manual and source code equations indicate that RELAP5/MOD3.3 should perform poorly in modelling interphase friction, actual RELAP5/MOD3.3 model runs perform very well in predicting pressure gradient, void fraction, the liquid and vapor velocities and the interphase friction force. This is largely due to RELAP5/MOD3.3 being able to adjust parameters to converge to a solution that fits within the boundary conditions established in the input file. Modifications to the RELAP5/MOD3.3 code were first made with the three correlations developed using dimensionless parameters, and were tested with data points that the RELAP5/MOD3.3 flow regime map had predicted would be annular flow. While the mixture Froude number correlation has been analyzed to be the most statistically accurate of the three correlations, it was found that the Weber number correlation performed best when implemented into RELAP5/MOD3.3. In a parametric study of the Weber number correlation, it performed optimally at 150% of the original correlation, improving upon the original RELAP model in almost every metric examined. Additional investigations were performed with individual annular flow correlations that model specific physical parameters. Results with the annular flow physical models were inconclusive as no particular model provided a significant improvement over the original RELAP5/MOD3.3 model, and there was no clear indication that combining the models would provide significant improvement. / Ph. D.

The purpose of this study is to develop an accurate model of the downcomer of the once-through steam generator (OTSG) developed by Babcock & Wilcox, using RELAP5/MOD2. While the physical model can be easily developed, several parameters are left to be adjusted to optimally model the downcomer and match data that was retrieved in a first-of-a-kind (FOAK) study conducted at Oconee Unit I in Oconee, South Carolina. Once the best-fit set of parameters has been determined, then the model must be tested for power levels exceeding that for which the steam generator was originally designed, so as to determine the power level at which a phenomenon known as flood-back becomes a concern. All known previous studies that have been conducted using RELAP5/MOD2 have shown that RELAP over-predicts interphase friction. However, all of those studies focused on heated two-phase upflow, whereas the downcomer is modeled as adiabatic two-phase downflow. In this study, it is found that the original slug drag model for RELAP5/MOD2 developed by Idaho National Engineering Laboratory (INEL) under-predicts the interphase friction between the liquid and vapor phase within the downcomer. Using a modified version of the original slug drag model created by Babcock & Wilcox (B&W), an optimum multiplier is found for each power level. An increase of 1181% in interphase friction over the INEL slug drag model, which equals an increase of 4347% for the default B&W model provides the most accurate results for all power levels studied. Emphasis is also placed on modeling the orifice plate of the OTSG downcomer which has been added to stabilize pressure fluctuations between the downcomer and tube bundle of the OTSG. While several different schemes are explored for modeling the orifice plate, a branch connection with an inlet area 14.22% of that of the downcomer is used to model the orifice plate along with the volume that transitions the two-phase downflow to horizontal flow into the tube nest of the OTSG. Power levels exceeding that for which the steam generator was designed are tested in RELAP using the slug drag multiplier to determine at which power level a liquid level would occur and would flood-back become a concern. In this study, it is determined that a liquid level would form at 135% power and that at any higher power level, flood-back would be of concern for any user of the steam generator. / Master of Science

Since the Fukushima Daiichi nuclear accident, there have been ongoing efforts to enhance the modelling capabilities for severe accidents in nuclear power plants. The primary severe accident analysis code used in Canada for its CANDU reactors is MAAP-CANDU (adapted from MAAP-LWR). In order to meet the new requirements that have evolved since Fukushima, upgrades to MAAP-CANDU have been made most recently by the Canadian nuclear industry. While the newest version (i.e. MAAP5-CANDU) offers several important improvements primarily in core nodalization and core collapse modelling, it still lacks mechanistic models for many key thermo-mechanical deformation phenomena that may significantly impact accident progression and event timings. It is also a general consensus that having alternative analysis tools is beneficial in improving our confidence in the simulation results, especially given the complex nature of severe accident phenomena in CANDU and the limited experimental support. This thesis seeks a novel approach to CANDU severe accident modelling by combining the best-estimate thermal-hydraulic code RELAP5, the severe accident models in SCDAP, and several CANDU-specific mechanistic deformation models developed by the author. This work mainly consists of two parts. The first part is focused on the assessment of natural circulation heat sinks following crash-cooldown in the early-phase of a Station Blackout (SBO) accident where fuel channel deformation can be precluded. The effectiveness of steam generator heat removal after crash-cooldown and that of the several water make-up options were demonstrated through the simulation of several SBO scenarios with/without crash-cooldown, sensitivity studies, as well as benchmarking against station and experimental measurements. In the second part, several mechanistic severe accident models were developed to enhance the simulation fidelity beyond the initial steam generator heat sink phase to the moderator boil-off and core disassembly phases. This includes models for predicting the pressure tube ballooning and sagging phenomena during the fuel channel heat-up phase and models for the sagging and disassembly of fuel channel assemblies during the core disassembly phase. After benchmarking against relevant channel deformation experiments, the models were successfully integrated into the RELAP/SCDAPSIM/MOD3.6 code as part of the SCDAP subroutines. The advantage of utilizing a code such as SCDAP is that generic models for fission product release and hydrogen generations, which are well benchmarked, can be directly applied to CANDU simulations. With the modified MOD3.6 code the early-phase SBO simulations were extended to include the later stages of SBO until the calandria vessel dryout. The current modelling approach replaced the simple threshold-type models commonly seen in the integrated severe accident codes such as MAAP-CANDU with more mechanistic models thereby providing a more robust treatment of the core degradation process during severe accident in CANDU. / Thesis / Doctor of Philosophy (PhD)

Im Rahmen eines vom BMBF/BMWi geförderten WTZ-Vorhabens wurden der Transfer des im Forschungszentrum Rossendorf (FZR) entwickelten Programmcodes DYN3D und seine Integration in die programmtechnische Basis des Instituts für Physik und Energietechnik (IPPE) Obninsk realisiert. Das neutronenkinetische Programmodul von DYN3D wurde von den russischen Experten genutzt, um den im IPPE verwendeten Thermohydraulikcode um die Funktion der 3D Neutronenkinetik zu erweitern. Zur Modernisierung der bisher bei Störfallanalysen verwendeten makroskopischen Wirkungsquerschnitte wurde mit dem Programmcode WIMS/D4 eine neue Datenbibliothek generiert, welche auch die bereits in WWER-Reaktoren umgesetzten Konzepte zu modifizierten Kernbeladungen (abbrennbare Absorber unterschiedlicher Borkonzentration) berücksichtigen kann. Diese Querschnittdatenbibliothek wurde an DYN3D angeschlossen und in ersten Vergleichsrechnungen auf seine Funktionstüchtigkeit sowie Genauigkeit überprüft. Für das unter Beteiligung von IPPE erstellte integrale Reaktorkonzept ABV-67 wurden sowohl mit DYN3D als auch mit dem gekoppelten Programmkomplex erste ATWS-Analysen durchgeführt. Der im IPPE entwickelte Fluiddynamikcode DINCOR wurde dem FZR zur Nutzung übergeben und durch gemeinsame Nachrechnungen von Modellaufgaben zum kurzzeitigen Schmelzeverhalten (CORVIS-Experimente) validiert.

Submitted by Almir Azevedo (barbio1313@gmail.com) on 2015-01-06T14:40:18Z No. of bitstreams: 1 dissertação mestrado ien 2014 Alexandre Soares.pdf: 4924692 bytes, checksum: 9690f1916310052f66cbc6e41f71c443 (MD5) / Made available in DSpace on 2015-01-06T14:40:18Z (GMT). No. of bitstreams: 1 dissertação mestrado ien 2014 Alexandre Soares.pdf: 4924692 bytes, checksum: 9690f1916310052f66cbc6e41f71c443 (MD5) Previous issue date: 2014-11 / O presente trabalho propões um estudo de um acidente com perda de refrigerante de um reator PWR através de um Simulador de Escopo Completo e do código computacional RELAP. Para tal, foi considerado um acidente com perda de refrigerante com área de quebra de 160 cm2 na perna fria do circuito 20 do sistema de refrigeração do reator da planta da Usina Nuclear de Angra 2, com o reator operando em condições estacionária, a 100% de potência. Foi admitido ainda, que ocorreu simultaneamente a perda de Suprimento Externo de Energia Elétrica e que a disponibilidade do Sistema de Refrigeração de Emergência do Núcleo não era plena. Os resultados obtidos apresentam-se bastante relevantes e com possibilidade de serem usados no planejamento de atividades futuras, visto que a construção de Angra 3 se apresenta em andamento e se assemelha a Angra 2. / The present paper porposes a study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP. To this end, it considered a loss of coolant accident with 160 cm2 breaking area in cold leg of 20 circuit of the reactor cooling system of nuclear power plant Angra 2, with the reactor operating in stationary condition, to 100% power. It considered that occurred at the same time the loss of External Power Supply and the availability of emergency cooling system was not full. The results obtained are quite relevant and with the possibility of being used in the planning of future activities, given that the construction of Angra 3 is underway and resembles the Angra 2.

Im Rahmen eines vom BMBF/BMWi geförderten WTZ-Vorhabens wurden der Transfer des im Forschungszentrum Rossendorf (FZR) entwickelten Programmcodes DYN3D und seine Integration in die programmtechnische Basis des Instituts für Physik und Energietechnik (IPPE) Obninsk realisiert. Das neutronenkinetische Programmodul von DYN3D wurde von den russischen Experten genutzt, um den im IPPE verwendeten Thermohydraulikcode um die Funktion der 3D Neutronenkinetik zu erweitern. Zur Modernisierung der bisher bei Störfallanalysen verwendeten makroskopischen Wirkungsquerschnitte wurde mit dem Programmcode WIMS/D4 eine neue Datenbibliothek generiert, welche auch die bereits in WWER-Reaktoren umgesetzten Konzepte zu modifizierten Kernbeladungen (abbrennbare Absorber unterschiedlicher Borkonzentration) berücksichtigen kann. Diese Querschnittdatenbibliothek wurde an DYN3D angeschlossen und in ersten Vergleichsrechnungen auf seine Funktionstüchtigkeit sowie Genauigkeit überprüft. Für das unter Beteiligung von IPPE erstellte integrale Reaktorkonzept ABV-67 wurden sowohl mit DYN3D als auch mit dem gekoppelten Programmkomplex erste ATWS-Analysen durchgeführt. Der im IPPE entwickelte Fluiddynamikcode DINCOR wurde dem FZR zur Nutzung übergeben und durch gemeinsame Nachrechnungen von Modellaufgaben zum kurzzeitigen Schmelzeverhalten (CORVIS-Experimente) validiert.

The safety performance of nuclear power reactors is a very important factor in evaluating nuclear energy sustainability. Improving the safety performance of nuclear reactors can enhance nuclear energy sustainability as it will improve the environmental indicator used to evaluate the overall sustainability of nuclear energy. Great interest is given now to advanced nuclear reactors especially those using passive safety components. Investigation of the improvement in nuclear safety using advanced reactors was done by comparing the safety performance of a conventional reactor which uses active safety systems, such as Pressurized Water Reactor (PWR), with an advanced reactor which uses passive safety systems, such as AP1000, during a design basis accident, such as Loss of Coolant Accident (LOCA), using the PCTran as a simulation code. To assess the safety performance of PWR and AP1000, the “Global Safety Index” GSI model was developed by introducing three indicators: probability of accident occurrence, performance of safety system in case of an accident occurrence, and the consequences of the accident. Only the second indicator was considered in this work. A more detailed model for studying the performance of passive safety systems in AP1000 was developed. That was done using SCDAPSIM/RELAP5 code as it is capable of modelling design basis accidents (DBAs) in advanced nuclear reactors.

[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 / TESIS

The nuclear industry has long relied upon bounding parametric analyses in predicting the safety margins of reactor designs undergoing design-basis accidents. These methods have been known to return highly-conservative results, limiting the operating conditions of the reactor. The Best-Estimate Plus Uncertainty (BEPU) method using a modernized version of the Code-Scaling, Applicability, and Uncertainty (CSAU) methodology has been applied to more accurately predict the safety margins of the Oregon State University Advanced Plant Experiment (APEX) facility experiencing a Loss-of-Feedwater Accident (LOFA). The statistical advantages of the Bayesian paradigm of probability was utilized to incorporate prior knowledge when determining the analysis required to justify the safety margins. RELAP5 Mod 3.3 was used to accurately predict the thermal-hydraulics of a primary Feed-and-Bleed response to the accident using assumptions to accompany the lumped-parameter calculation approach. A novel coupling of thermal-hydraulic and statistical software was accomplished using the Symbolic Nuclear Analysis Package (SNAP). Uncertainty in Peak Cladding Temperature (PCT) was calculated at the 95/95 probability/confidence levels under a series of four separate sensitivity studies. / Graduation date: 2013

The study and modelling of two-phase flow, even the simplest ones such as the bubbly flow, remains a challenge that requires exploring the physical phenomena from different spatial and temporal resolution levels. CFD (Computational Fluid Dynamics) is a widespread and promising tool for modelling, but nowadays, there is no single approach or method to predict the dynamics of these systems at the different resolution levels providing enough precision of the results. The inherent difficulties of the events occurring in this flow, mainly those related with the interface between phases, makes that low or intermediate resolution level approaches as system codes (RELAP, TRACE, ...) or 3D TFM (Two-Fluid Model) have significant issues to reproduce acceptable results, unless well-known scenarios and global values are considered. Instead, methods based on high resolution level such as Interfacial Tracking Method (ITM) or Volume Of Fluid (VOF) require a high computational effort that makes unfeasible its use in complex systems. In this thesis, an open-source simulation framework has been designed and developed using the OpenFOAM library to analyze the cases from microescale to macroscale levels. The different approaches and the information that is required in each one of them have been studied for bubbly flow. In the first part, the dynamics of single bubbles at a high resolution level have been examined through VOF. This technique has allowed to obtain accurate results related to the bubble formation, terminal velocity, path, wake and instabilities produced by the wake. However, this approach has been impractical for real scenarios with more than dozens of bubbles. Alternatively, this thesis proposes a CFD Discrete Element Method (CFD-DEM) technique, where each bubble is represented discretely. A novel solver for bubbly flow has been developed in this thesis. This includes a large number of improvements necessary to reproduce the bubble-bubble and bubble-wall interactions, turbulence, velocity seen by the bubbles, momentum and mass exchange term over the cells or bubble expansion, among others. But also new implementations as an algorithm to seed the bubbles in the system have been incorporated. As a result, this new solver gives more accurate results as the provided up to date. Following the decrease on resolution level, and therefore the required computational resources, a 3D TFM have been developed with a population balance equation solved with an implementation of the Quadrature Method Of Moments (QMOM). The solver is implemented with the same closure models as the CFD-DEM to analyze the effects involved with the lost of information due to the averaging of the instantaneous Navier-Stokes equation. The analysis of the results with CFD-DEM reveals the discrepancies found by considering averaged values and homogeneous flow in the models of the classical TFM formulation. Finally, for the lowest resolution level approach, the system code RELAP5/MOD3 is used for modelling the bubbly flow regime. The code has been modified to reproduce properly the two-phase flow characteristics in vertical pipes, comparing the performance of the calculation of the drag term based on drift-velocity and drag coefficient approaches. / El estudio y modelado de flujos bifásicos, incluso los más simples como el bubbly flow, sigue siendo un reto que conlleva aproximarse a los fenómenos físicos que lo rigen desde diferentes niveles de resolución espacial y temporal. El uso de códigos CFD (Computational Fluid Dynamics) como herramienta de modelado está muy extendida y resulta prometedora, pero hoy por hoy, no existe una única aproximación o técnica de resolución que permita predecir la dinámica de estos sistemas en los diferentes niveles de resolución, y que ofrezca suficiente precisión en sus resultados. La dificultad intrínseca de los fenómenos que allí ocurren, sobre todo los ligados a la interfase entre ambas fases, hace que los códigos de bajo o medio nivel de resolución, como pueden ser los códigos de sistema (RELAP, TRACE, etc.) o los basados en aproximaciones 3D TFM (Two-Fluid Model) tengan serios problemas para ofrecer resultados aceptables, a no ser que se trate de escenarios muy conocidos y se busquen resultados globales. En cambio, códigos basados en alto nivel de resolución, como los que utilizan VOF (Volume Of Fluid), requirieren de un esfuerzo computacional tan elevado que no pueden ser aplicados a sistemas complejos. En esta tesis, mediante el uso de la librería OpenFOAM se ha creado un marco de simulación de código abierto para analizar los escenarios desde niveles de resolución de microescala a macroescala, analizando las diferentes aproximaciones, así como la información que es necesaria aportar en cada una de ellas, para el estudio del régimen de bubbly flow. En la primera parte se estudia la dinámica de burbujas individuales a un alto nivel de resolución mediante el uso del método VOF (Volume Of Fluid). Esta técnica ha permitido obtener resultados precisos como la formación de la burbuja, velocidad terminal, camino recorrido, estela producida por la burbuja e inestabilidades que produce en su camino. Pero esta aproximación resulta inviable para entornos reales con la participación de más de unas pocas decenas de burbujas. Como alternativa, se propone el uso de técnicas CFD-DEM (Discrete Element Methods) en la que se representa a las burbujas como partículas discretas. En esta tesis se ha desarrollado un nuevo solver para bubbly flow en el que se han añadido un gran número de nuevos modelos, como los necesarios para contemplar los choques entre burbujas o con las paredes, la turbulencia, la velocidad vista por las burbujas, la distribución del intercambio de momento y masas con el fluido en las diferentes celdas por cada una de las burbujas o la expansión de la fase gaseosa entre otros. Pero también se han tenido que incluir nuevos algoritmos como el necesario para inyectar de forma adecuada la fase gaseosa en el sistema. Este nuevo solver ofrece resultados con un nivel de resolución superior a los desarrollados hasta la fecha. Siguiendo con la reducción del nivel de resolución, y por tanto los recursos computacionales necesarios, se efectúa el desarrollo de un solver tridimensional de TFM en el que se ha implementado el método QMOM (Quadrature Method Of Moments) para resolver la ecuación de balance poblacional. El solver se desarrolla con los mismos modelos de cierre que el CFD-DEM para analizar los efectos relacionados con la pérdida de información debido al promediado de las ecuaciones instantáneas de Navier-Stokes. El análisis de resultados de CFD-DEM permite determinar las discrepancias encontradas por considerar los valores promediados y el flujo homogéneo de los modelos clásicos de TFM. Por último, como aproximación de nivel de resolución más bajo, se investiga el uso uso de códigos de sistema, utilizando el código RELAP5/MOD3 para analizar el modelado del flujo en condiciones de bubbly flow. El código es modificado para reproducir correctamente el flujo bifásico en tuberías verticales, comparando el comportamiento de aproximaciones para el cálculo del término d / L'estudi i modelatge de fluxos bifàsics, fins i tot els més simples com bubbly flow, segueix sent un repte que comporta aproximar-se als fenòmens físics que ho regeixen des de diferents nivells de resolució espacial i temporal. L'ús de codis CFD (Computational Fluid Dynamics) com a eina de modelatge està molt estesa i resulta prometedora, però ara per ara, no existeix una única aproximació o tècnica de resolució que permeta predir la dinàmica d'aquests sistemes en els diferents nivells de resolució, i que oferisca suficient precisió en els seus resultats. Les dificultat intrínseques dels fenòmens que allí ocorren, sobre tots els lligats a la interfase entre les dues fases, fa que els codis de baix o mig nivell de resolució, com poden ser els codis de sistema (RELAP,TRACE, etc.) o els basats en aproximacions 3D TFM (Two-Fluid Model) tinguen seriosos problemes per a oferir resultats acceptables , llevat que es tracte d'escenaris molt coneguts i se persegueixen resultats globals. En canvi, codis basats en alt nivell de resolució, com els que utilitzen VOF (Volume Of Fluid), requereixen d'un esforç computacional tan elevat que no poden ser aplicats a sistemes complexos. En aquesta tesi, mitjançant l'ús de la llibreria OpenFOAM s'ha creat un marc de simulació de codi obert per a analitzar els escenaris des de nivells de resolució de microescala a macroescala, analitzant les diferents aproximacions, així com la informació que és necessària aportar en cadascuna d'elles, per a l'estudi del règim de bubbly flow. En la primera part s'estudia la dinàmica de bambolles individuals a un alt nivell de resolució mitjançant l'ús del mètode VOF. Aquesta tècnica ha permès obtenir resultats precisos com la formació de la bambolla, velocitat terminal, camí recorregut, estela produida per la bambolla i inestabilitats que produeix en el seu camí. Però aquesta aproximació resulta inviable per a entorns reals amb la participació de més d'unes poques desenes de bambolles. Com a alternativa en aqueix cas es proposa l'ús de tècniques CFD-DEM (Discrete Element Methods) en la qual es representa a les bambolles com a partícules discretes. En aquesta tesi s'ha desenvolupat un nou solver per a bubbly flow en el qual s'han afegit un gran nombre de nous models, com els necessaris per a contemplar els xocs entre bambolles o amb les parets, la turbulència, la velocitat vista per les bambolles, la distribució de l'intercanvi de moment i masses amb el fluid en les diferents cel·les per cadascuna de les bambolles o els models d'expansió de la fase gasosa entre uns altres. Però també s'ha hagut d'incloure nous algoritmes com el necessari per a injectar de forma adequada la fase gasosa en el sistema. Aquest nou solver ofereix resultats amb un nivell de resolució superior als desenvolupat fins la data. Seguint amb la reducció del nivell de resolució, i per tant els recursos computacionals necessaris, s'efectua el desenvolupament d'un solver tridimensional de TFM en el qual s'ha implementat el mètode QMOM (Quadrature Method Of Moments) per a resoldre l'equació de balanç poblacional. El solver es desenvolupa amb els mateixos models de tancament que el CFD-DEM per a analitzar els efectes relacionats amb la pèrdua d'informació a causa del promitjat de les equacions instantànies de Navier-Stokes. L'anàlisi de resultats de CFD-DEM permet determinar les discrepàncies ocasionades per considerar els valors promitjats i el flux homogeni dels models clàssics de TFM. Finalment, com a aproximació de nivell de resolució més baix, s'analitza l'ús de codis de sistema, utilitzant el codi RELAP5/MOD3 per a analitzar el modelatge del fluxos en règim de bubbly flow. El codi és modificat per a reproduir correctament les característiques del flux bifàsic en canonades verticals, comparant el comportament d'aproximacions per al càlcul del terme de drag basades en velocitat de drift flux model i de les basades en coe / Peña Monferrer, C. (2017). Computational fluid dynamics multiscale modelling of bubbly flow. A critical study and new developments on volume of fluid, discrete element and two-fluid methods [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/90493 / TESIS