[ES] La simulación de transitorios forma parte del proceso de licenciamiento de una central nuclear. Esto implica que los códigos, así como los modelos utilizados deben estar verificados y validados. Normalmente, esta simulación se realiza con códigos termohidráulicos de planta que tienen una definición de la cinética del reactor muy simplificada con cinética puntual o unidimensional. Una mejora importante en la simulación de transitorios base de diseño se basa en la utilización de códigos acoplados termohidráulico-neutrónicos, que permiten obtener resultados sobre la evolución de la potencia del reactor en tres dimensiones. Los códigos neutrónicos 3D necesitan parámetros de la cinética y secciones eficaces también en 3D ajustados al punto del ciclo que se quiere simular y que abarquen las condiciones que se alcancen durante el transitorio. Por otro lado, para poder verificar tanto los códigos como los modelos es necesario llevar a cabo una serie de simulaciones de diferentes transitorios. De esta manera, se comprueba cómo funciona el código acoplado en diferentes condiciones de operación y simulación. Esta tesis contribuye al conocimiento del uso de códigos termohidráulico-neutrónicos acoplados en la simulación de transitorios base de diseño (Design Basis Accidents -DBAs). Los códigos mejorados y verificados son los códigos termohidráulicos RELAP5, TRAC-BF1 y TRACE y el código neutrónico PARCS. Los parámetros neutrónicos necesarios en PARCS se han obtenido aplicando una metodología que simplifica el modelo del núcleo. Esta metodología, ya desarrollada e implementada, denominada SIMTAB, se ha mejorado, tanto en las posibilidades de aplicación de la misma como en la optimización y actualización de la programación del código fuente. Los transitorios analizados con los códigos RELAP5/PARCS acoplados son: transitorio por expulsión de barra de control y transitorio de inyección de boro en un reactor PWR. Con los códigos TRAC-BF1/PARCS acoplados se ha analizado el transitorio por disparo de turbina en la C. N. Peach Bottom. Para llevar a cabo las simulaciones con TRAC-BF1/PARCS se ha implementado el acoplamiento de ambos códigos, puesto que originalmente el código TRAC-BF1 no estaba preparado para ello. El análisis de inestabilidades en reactores BWR se ha realizado con RELAP5/PARCS en dos reactores BWR: C. N. Peach Bottom y C. N. Ringhals 1. Para ello se ha desarrollado una metodología de análisis que abarca desde la definición del modelo termohidráulico y del modelo neutrónico hasta el análisis de las señales simuladas obtenidas con PARCS. La metodología también incluye la aplicación de diferentes perturbaciones basadas en los modos Lambda y en el análisis de las señales reales de planta. Se ha llevado a cabo un estudio del modelo para el cálculo de la concentración de Boro en los códigos termohidráulicos y se ha mejorado este modelo en el código TRAC-BF1, incorporando un nuevo método de resolución en el código fuente. El modelo para el cálculo del calor de desintegración también se ha revisado y mejorado en los códigos TRAC-BF1 y PARCS. En ambos casos se ha implementado el modelo ANS 2005. El análisis de sensibilidad e incertidumbre está ligado a los resultados de los códigos de mejor estimación como los mejorados en esta tesis. Este análisis se ha realizado sobre los transitorios de expulsión de barra en un reactor PWR y el transitorio de caída de barra en un reactor BWR con RELAP5/PARCS. Los resultados de estos trabajos aportan una metodología de aplicación para la simulación correcta de transitorios con códigos acoplados. Además, ha servido para detectar y subsanar deficiencias en los códigos, y de esta manera disponer de unos códigos de mejor estimación preparados para el análisis de transitorios base de diseño. / [CA] La simulació de transitoris forma part del procés de llicenciament d'una central nuclear. Això implica que els codis, així com els models utilitzats han d'estar verificats i validats. Normalment, aquesta simulació es realitza amb codis termohidràulics de planta que tenen una definició de la cinètica del reactor molt simplificada amb cinètica puntual o unidimensional. Una millora important en la simulació de transitoris base de disseny es basa en la utilització de codis acoblats termohidràulic-neutrònics, que permeten obtindre resultats sobre l'evolució de la potència del reactor en tres dimensions. Els codis neutrònics 3D necessiten paràmetres de la cinètica i seccions eficaces també en 3D ajustats al punt del cicle que es vol simular i que abasten les condicions que s'aconseguisquen durant el transitori. D'altra banda, per a poder verificar tant els codis com els models és necessari dur a terme una sèrie de simulacions de diferents transitoris. D'aquesta manera, es comprova com funciona el codi acoblat en diferents condicions d'operació i simulació. Aquesta tesi contribueix al coneixement de l'ús de codis termohidràulic-neutrònics acoblats en la simulació de transitoris base de disseny. Els codis millorats i verificats són els codis termohidràulics RELAP5, TRAC-BF1 i TRACE i el codi neutrònic PARCS. Els paràmetres neutrònics necessaris en PARCS s'han obtingut aplicant una metodologia que simplifica el model del nucli. Aquesta metodologia, ja desenvolupada i implementada, denominada SIMTAB, s'ha millorat, tant en les possibilitats d'aplicació de la mateixa com en l'optimització i actualització de la programació del codi font. Els transitoris analitzats amb els codis RELAP5/PARCS acoblats són: transitori per expulsió de barra de control i transitori d'injecció de bor en un reactor PWR. Amb els codis TRAC-BF1/PARCS acoblats s'ha analitzat el transitori per disparament de turbina en la C. N. Peach Bottom. Per a dur a terme les simulacions amb TRAC-BF1/PARCS s'ha implementat l'acoblament de tots dos codis, ja que originalment el codi TRAC-BF1 no estava preparat per a això. L'anàlisi d'inestabilitats en reactors BWR s'ha realitzat amb RELAP5/PARCS en dos reactors BWR: C. N. Peach Bottom i C. N. Ringhals 1. Per a això s'ha desenvolupat una metodologia d'anàlisi que abasta des de la definició del model termohidràulic i del model neutrònic fins a l'anàlisi dels senyals simulats. La metodologia també inclou l'aplicació de diferents pertorbacions basades en els modes Lambda i en l'anàlisi dels senyals reals de planta. S'ha dut a terme un estudi del model per al càlcul de la concentració de Bor en els codis termohidràulics i s'ha millorat aquest model en el codi TRAC-BF1, incorporant un nou mètode de resolució en el codi font. El model per al càlcul de la calor de desintegració també s'ha revisat i millorat en els codis TRAC-BF1 i PARCS. En tots dos casos s'ha implementat el model ANS 2005. L'anàlisi de sensibilitat i incertesa està lligat als resultats dels codis de millor estimació com els millorats en aquesta tesi. Aquesta anàlisi s'ha realitzat sobre els transitoris d'expulsió de barra en un reactor PWR i el transitori de caiguda de barra en un reactor BWR amb RELAP5/PARCS. Els resultats d'aquests treballs aporten una metodologia d'aplicació per a la simulació correcta de transitoris amb codis acoblats. A més, ha servit per a detectar i esmenar deficiències en els codis, i d'aquesta manera disposar d'uns codis de millor estimació preparats per a l'anàlisi de transitoris base de disseny. / [EN] The simulation of transients is part of the licensing process of a nuclear power plant. This implies that the codes as well as the models used must be verified and validated. Normally, this simulation is performed with thermalhydraulic plant codes that have a very simplified definition of reactor kinetics with point or one-dimensional kinetics. An important improvement in the simulation of design-basis transients rely on the use of thermohydraulic-neutronic coupled codes, which allow to obtain results of the evolution of the reactor power in three dimensions. The 3D neutron codes need parameters of the kinetics and cross-sections also in 3D adjusted to the point of the cycle to be simulated that must cover the conditions reached during the transient. On the other hand, to be able to verify both the codes and the models it is necessary to carry out a series of simulations of different transients. In this way, it is checked how the coupled code works in different operating and simulation conditions. This thesis contributes to increase the knowledge of the use of thermalhydraulic-neutronic coupled codes in the simulation of design basis accidents (DBAs). The improved and verified codes are the thermalhydraulic codes RELAP5, TRAC-BF1 and TRACE and the neutronic code PARCS. The necessary neutronic parameters in PARCS have been obtained by applying a methodology that simplifies the core model. This methodology, already developed and implemented, called SIMTAB, has been improved in this thesis in its application possibilities and also in the optimization and updating of the source code. The transients analyzed with RELAP5/PARCS coupled code are: control rod ejection transient and boron injection transient in a PWR reactor. With TRAC-BF1/PARCS coupled code, the transient analyzed is the turbine trip transient in Peach Bottom NPP. To carry out the simulations with TRAC-BF1/PARCS, the coupling of both codes has been implemented before, since originally the TRAC-BF1 code was not prepared for it. The analysis of instabilities in BWR reactors has been carried out with RELAP5/PARCS in two BWR reactors: Peach Bottom NPP and Ringhals 1 NPP. A methodology has been developed which cover from the definition of the thermalhydraulic model and the neutron model to the simulated signal analysis. The methodology also includes the application of different disturbances based on Lambda modes and the analysis of real plant signals. A study of the model for the calculation of the Boron concentration in thermalhydraulic codes has been carried out. This model has been improved in the TRAC-BF1 code, incorporating a new resolution method in the source code. The model for the calculation of the decay heat has also been revised and improved in TRAC-BF1 and PARCS codes. In both cases, the ANS 2005 model has been implemented. The sensitivity and uncertainty analysis is linked to the results of the best estimate codes such as those improved in this thesis. This analysis has been carried out on the control rod ejection transients in a PWR reactor and the control rod drop transient in a BWR reactor with RELAP5/PARCS. The results of these works provide an application methodology for the correct simulation of transients with coupled codes. In addition, it has been used to detect and correct deficiencies in the codes, and therefore, to have better estimate codes prepared for the analysis of design-basis transients. / Barrachina Celda, TM. (2020). Aportaciones y Mejoras en los Códigos Termohidráulicos y Neutrónicos de Estimación Óptima RELAP5, TRAC-BF1, TRACE Y PARCS [Tesis doctoral]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/158745 / TESIS

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.

The Pebble Bed Modular Reactor (PBMR) is one of the most technologically advanced developments in South Africa. In order to build a commercially viable demonstration power plant, all the specifically and uniquely designed equipment must first be qualified. All the prototype equipment is tested at the Helium Test Facility (HTF) at Pelindaba. One of the largest components that are tested is the Core Unloading Device (CUD). The main function of the CUD is to unload fuel from the bottom of the reactor core to enable circulation of the fuel core. The CUD housing vessel forms part of the reactor pressure boundary. Pebble-directing valves and other moving machinery are installed inside its machined inner surface. It is essential that the interior surfaces of the CUD are case hardened to provide a corrosion- and wear-resistant layer. Cold welding between the moving metal parts and the machined surface must also be prevented. Nitriding is a case hardening process that adds a hardened wear- and corrosion-resistant layer that will also prevent cold welding of the moving parts in the helium atmosphere. Only a few nitriding furnaces exist that can house a forging as large as the CUD of the PBMR. Commercial nitriding furnaces in South Africa are all too small and have limited flexibility in terms of the nitriding process. The nitriding of a vessel as large as the CUD has not yet been carried out commercially. The aim of this work was to design and develop a custom-made nitriding plant to perform the nitriding of the first PBMR/HTF CUD. Proper process control is essential to ensure that the required nitrided case has been obtained. A new concept for a gas nitriding plant was developed using the nitrided vessel interior as the nitriding process chamber. Before the commencement of detail design, a laboratory test was performed on a scale model vessel to confirm concept feasibility. The design of the plant included the mechanical design of various components essential to the nitriding process. A special stirring fan with an extended length shaft was designed, taking whirling speed into account. Considerable research was performed on the high temperature use of the various components to ensure the safe operation of the plant at temperatures of up to 600°C. Nitriding requires the use of hazardous gases such as ammonia, oxygen and nitrogen. Hydrogen is produced as a by-product and therefore safety was the most important design parameter. Thermohydraulic analyses, i.e. heat transfer and pressure drop calculations in pipes, were also performed to ensure the successful process design of the nitriding plant. The nitriding plant was subsequently constructed and operated to verify the correct design. A large amount of experimental and operating data was captured during the actual operation of the plant. This data was analysed and the thermohydraulic analyses were verified. Nitrided specimens were subjected to hardness and layer thickness tests. The measured temperature of the protruding fan shaft was within the limits predicted by Finite Element Analysis (FEA) models. Graphs of gas flow rates and other operation data confirmed the inverse proportionality between ammonia supply flow rate and measured dissociation rate. The design and operation of the nitriding plant were successful as a nitride layer thickness of 400 μm and hardness of 1 200 Vickers hardness (VHN) was achieved. This research proves that a large pressure vessel can successfully be nitrided using the vessel interior as a process chamber. / Thesis (M.Ing. (Mechanical Engineering))--North-West University, Potchefstroom Campus, 2010.

The Pebble Bed Modular Reactor (PBMR) is one of the most technologically advanced developments in South Africa. In order to build a commercially viable demonstration power plant, all the specifically and uniquely designed equipment must first be qualified. All the prototype equipment is tested at the Helium Test Facility (HTF) at Pelindaba. One of the largest components that are tested is the Core Unloading Device (CUD). The main function of the CUD is to unload fuel from the bottom of the reactor core to enable circulation of the fuel core. The CUD housing vessel forms part of the reactor pressure boundary. Pebble-directing valves and other moving machinery are installed inside its machined inner surface. It is essential that the interior surfaces of the CUD are case hardened to provide a corrosion- and wear-resistant layer. Cold welding between the moving metal parts and the machined surface must also be prevented. Nitriding is a case hardening process that adds a hardened wear- and corrosion-resistant layer that will also prevent cold welding of the moving parts in the helium atmosphere. Only a few nitriding furnaces exist that can house a forging as large as the CUD of the PBMR. Commercial nitriding furnaces in South Africa are all too small and have limited flexibility in terms of the nitriding process. The nitriding of a vessel as large as the CUD has not yet been carried out commercially. The aim of this work was to design and develop a custom-made nitriding plant to perform the nitriding of the first PBMR/HTF CUD. Proper process control is essential to ensure that the required nitrided case has been obtained. A new concept for a gas nitriding plant was developed using the nitrided vessel interior as the nitriding process chamber. Before the commencement of detail design, a laboratory test was performed on a scale model vessel to confirm concept feasibility. The design of the plant included the mechanical design of various components essential to the nitriding process. A special stirring fan with an extended length shaft was designed, taking whirling speed into account. Considerable research was performed on the high temperature use of the various components to ensure the safe operation of the plant at temperatures of up to 600°C. Nitriding requires the use of hazardous gases such as ammonia, oxygen and nitrogen. Hydrogen is produced as a by-product and therefore safety was the most important design parameter. Thermohydraulic analyses, i.e. heat transfer and pressure drop calculations in pipes, were also performed to ensure the successful process design of the nitriding plant. The nitriding plant was subsequently constructed and operated to verify the correct design. A large amount of experimental and operating data was captured during the actual operation of the plant. This data was analysed and the thermohydraulic analyses were verified. Nitrided specimens were subjected to hardness and layer thickness tests. The measured temperature of the protruding fan shaft was within the limits predicted by Finite Element Analysis (FEA) models. Graphs of gas flow rates and other operation data confirmed the inverse proportionality between ammonia supply flow rate and measured dissociation rate. The design and operation of the nitriding plant were successful as a nitride layer thickness of 400 μm and hardness of 1 200 Vickers hardness (VHN) was achieved. This research proves that a large pressure vessel can successfully be nitrided using the vessel interior as a process chamber. / Thesis (M.Ing. (Mechanical Engineering))--North-West University, Potchefstroom Campus, 2010.