El Generation IV International Forum (GIF) [1] es un programa internacional dedicado a apoyar, coordinar y dirigir las iniciativas de investigación y desarrollo encaminados a implementar las soluciones tecnológicas que caracterizarán a la siguiente generación de reactores nucleares. Estos reactores se caracterizaran por una gestión más eficiente del combustible nuclear, un incremento en las exigencias de seguridad y una alta competitividad económica. Con tales objetivos, GIF propuso una serie de diseños potencialmente capaces de alcanzarlos. Estos diseños son tecnológicamente muy distintos a las plantas nucleares comerciales actuales al utilizar neutrones de espectro rápido y consecuentemente refrigeración por metales líquidos. Estos nuevos diseños requieren el desarrollo y validación de herramientas computacionales capaces de simular el comportamiento de la planta tanto en fase estacionaria como en transitoria y por tanto sean aplicables en los procesos de diseño y licitación de dichas plantas. El objetivo de esta tesis es el de adaptar los códigos computacionales actuales aplicados a la simulación de reactores refrigerados por agua a reactores rápidos refrigerados por metales líquidos, tales como el sodio o el plomo y el desarrollo de modelos capaces de simular de una manera consistente el comportamientos de los sistemas ante determinados eventos que constituyen la base de diseño de la planta Para ello se adaptaran dichos códigos a la fenomenología específica de estos reactores, se desarrollaran modelos termo-hidráulicos y neutrónicos tanto unidimensionales como tridimensionales de los diseños propuestos y se validarán los resultados para demostrar su aplicabilidad. El trabajo incluye la implementación de correlaciones específicas para habilitar los códigos para el cálculo de la condiciones termo-hidráulicas de los refrigerantes así como la adaptación de los esquemas de acoplamiento termo-hidráulico-neutrónicos existentes a esta nueva tecnología. / Lázaro Chueca, A. (2014). Development, assessment and application of computational tools for design safety analysis of liquid metal cooled fast breeder reactors [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/39353 / TESIS

International Atomic Energy Agency (IAEA) safeguards are one method of proliferation resistance which is applied at most nuclear facilities worldwide. IAEA safeguards act to prevent the diversion of nuclear materials from a facility through the deterrence of detection. However, even with IAEA safeguards present at a facility, the country where the facility is located may still attempt to proliferate nuclear material by exploiting weaknesses in the safeguards system. The IAEA's mission is to detect the diversion of nuclear materials as soon as possible and ideally before it can be weaponized. Modern IAEA safeguards utilize unattended monitoring systems (UMS) to perform nuclear material accountancy and maintain the continuity of knowledge with regards to the position of nuclear material at a facility. This research focuses on evaluating the reliability of unattended monitoring systems and integrating the probabilistic failure of these systems into the comprehensive probabilistic proliferation resistance model of a facility. To accomplish this, this research applies reliability engineering analysis methods to probabilistic proliferation resistance modeling. This approach is demonstrated through the analysis of a safeguards design for the Example Sodium Fast Reactor Fuel Cycle Facility (ESFR FCF). The ESFR FCF UMS were analyzed to demonstrate the analysis and design processes that an analyst or designer would go through when evaluating/designing the proliferation resistance component of a safeguards system. When comparing the mean time to failure (MTTF) for the system without redundancies versus one with redundancies, it is apparent that redundancies are necessary to achieve a design without routine failures. A reliability engineering approach to probabilistic safeguards system analysis and design can be used to reach meaningful conclusions regarding the proliferation resistance of a UMS. The methods developed in this research provide analysts and designers alike a process to follow to evaluate the reliability of a UMS.

This master thesis deals steam generator for ESFR (European Sodium Fast Reactor), which is heated by liquid sodium. In the beginning chapters, there are theoretic information about ESFR's parameters and its' comparison with ohter types of heat exchangers in nuclear reactors with the same principal (sodium as a coolant). Then designing part follows, which contents of introduction of calculations, option of material and conception of heater. Computational part on its own includes thermal, hydraulic and stress calculations and comparison with aspects in nuclear safety and security.

The thesis is mainly focused on the design of the intermediate heat exchanger primary circuit of the reactor ESFR. Heat exchanger is calculated heat, hydraulic and strength and is finally processed part drawings. There are designed the basic dimensions of the tube bundle and container heat exchanger. There are included an overview of concepts and so far used types IHX at the nuclear power plants with fast reactors. There are also mentioned basic parameters of the project ESFR and evaluated the safety and operational reliability of the heat exchanger.