Diseño basado en técnicas de fiabilidad del tratamiento de mejora del terreno mediante columnas de grava

El proyecto geotécnico de columnas de grava tiene todas las incertidumbres asociadas a un proyecto geotécnico y además hay que considerar las incertidumbres inherentes a la compleja interacción entre el terreno y la columna, la puesta en obra de los materiales y el producto final conseguido. Este hecho es común a otros tratamientos del terreno cuyo objetivo sea, en general, la mejora “profunda”. Como los métodos de fiabilidad (v.gr., FORM, SORM, Monte Carlo, Simulación Direccional) dan respuesta a la incertidumbre de forma mucho más consistente y racional que el coeficiente de seguridad tradicional, ha surgido un interés reciente en la aplicación de técnicas de fiabilidad a la ingeniería geotécnica. Si bien la aplicación concreta al proyecto de técnicas de mejora del terreno no es tan extensa. En esta Tesis se han aplicado las técnicas de fiabilidad a algunos aspectos del proyecto de columnas de grava (estimación de asientos, tiempos de consolidación y aumento de la capacidad portante) con el objetivo de efectuar un análisis racional del proceso de diseño, considerando los efectos que tienen la incertidumbre y la variabilidad en la seguridad del proyecto, es decir, en la probabilidad de fallo. Para alcanzar este objetivo se ha utilizado un método analítico avanzado debido a Castro y Sagaseta (2009), que mejora notablemente la predicción de las variables involucradas en el diseño del tratamiento y su evolución temporal (consolidación). Se ha estudiado el problema del asiento (valor y tiempo de consolidación) en el contexto de la incertidumbre, analizando dos modos de fallo: i) el primer modo representa la situación en la que es posible finalizar la consolidación primaria, parcial o totalmente, del terreno mejorado antes de la ejecución de la estructura final, bien sea por un precarga o porque la carga se pueda aplicar gradualmente sin afectar a la estructura o instalación; y ii) por otra parte, el segundo modo de fallo implica que el terreno mejorado se carga desde el instante inicial con la estructura definitiva o instalación y se comprueba que el asiento final (transcurrida la consolidación primaria) sea lo suficientemente pequeño para que pueda considerarse admisible. Para trabajar con valores realistas de los parámetros geotécnicos, los datos se han obtenido de un terreno real mejorado con columnas de grava, consiguiendo, de esta forma, un análisis de fiabilidad más riguroso. La conclusión más importante, obtenida del análisis de este caso particular, es la necesidad de precargar el terreno mejorado con columnas de grava para conseguir que el asiento ocurra de forma anticipada antes de la aplicación de la carga correspondiente a la estructura definitiva. De otra forma la probabilidad de fallo es muy alta, incluso cuando el margen de seguridad determinista pudiera ser suficiente. En lo que respecta a la capacidad portante de las columnas, existen un buen número de métodos de cálculo y de ensayos de carga (tanto de campo como de laboratorio) que dan predicciones dispares del valor de la capacidad última de las columnas de grava. En las mallas indefinidas de columnas, los resultados del análisis de fiabilidad han confirmado las consideraciones teóricas y experimentales existentes relativas a que no se produce fallo por estabilidad, obteniéndose una probabilidad de fallo prácticamente nula para este modo de fallo. Sin embargo, cuando se analiza, en el contexto de la incertidumbre, la capacidad portante de pequeños grupos de columnas bajo zapatas se ha obtenido, para un caso con unos parámetros geotécnicos típicos, que la probabilidad de fallo es bastante alta, por encima de los umbrales normalmente admitidos para Estados Límite Últimos. Por último, el trabajo de recopilación sobre los métodos de cálculo y de ensayos de carga sobre la columna aislada ha permitido generar una base de datos suficientemente amplia como para abordar una actualización bayesiana de los métodos de cálculo de la columna de grava aislada. El marco bayesiano de actualización ha resultado de utilidad en la mejora de las predicciones de la capacidad última de carga de la columna, permitiendo “actualizar” los parámetros del modelo de cálculo a medida que se dispongan de ensayos de carga adicionales para un proyecto específico. Constituye una herramienta valiosa para la toma de decisiones en condiciones de incertidumbre ya que permite comparar el coste de los ensayos adicionales con el coste de una posible rotura y , en consecuencia, decidir si es procedente efectuar dichos ensayos. The geotechnical design of stone columns has all the uncertainties associated with a geotechnical project and those inherent to the complex interaction between the soil and the column, the installation of the materials and the characteristics of the final (as built) column must be considered. This is common to other soil treatments aimed, in general, to “deep” soil improvement. Since reliability methods (eg, FORM, SORM, Monte Carlo, Directional Simulation) deals with uncertainty in a much more consistent and rational way than the traditional safety factor, recent interest has arisen in the application of reliability techniques to geotechnical engineering. But the specific application of these techniques to soil improvement projects is not as extensive. In this thesis reliability techniques have been applied to some aspects of stone columns design (estimated settlements, consolidation times and increased bearing capacity) to make a rational analysis of the design process, considering the effects of uncertainty and variability on the safety of the project, i.e., on the probability of failure. To achieve this goal an advanced analytical method due to Castro and Sagaseta (2009), that significantly improves the prediction of the variables involved in the design of treatment and its temporal evolution (consolidation), has been employed. This thesis studies the problem of stone column settlement (amount and speed) in the context of uncertainty, analyzing two failure modes: i) the first mode represents the situation in which it is possible to cause primary consolidation, partial or total, of the improved ground prior to implementation of the final structure, either by a pre-load or because the load can be applied gradually or programmed without affecting the structure or installation; and ii) on the other hand, the second mode implies that the improved ground is loaded from the initial instant with the final structure or installation, expecting that the final settlement (elapsed primary consolidation) is small enough to be allowable. To work with realistic values of geotechnical parameters, data were obtained from a real soil improved with stone columns, hence producing a more rigorous reliability analysis. The most important conclusion obtained from the analysis of this particular case is the need to preload the stone columns-improved soil to make the settlement to occur before the application of the load corresponding to the final structure. Otherwise the probability of failure is very high, even when the deterministic safety margin would be sufficient. With respect to the bearing capacity of the columns, there are numerous methods of calculation and load tests (both for the field and the laboratory) giving different predictions of the ultimate capacity of stone columns. For indefinite columns grids, the results of reliability analysis confirmed the existing theoretical and experimental considerations that no failure occurs due to the stability failure mode, therefore resulting in a negligible probability of failure. However, when analyzed in the context of uncertainty (for a case with typical geotechnical parameters), results show that the probability of failure due to the bearing capacity failure mode of a group of columns is quite high, above thresholds usually admitted for Ultimate Limit States. Finally, the review of calculation methods and load tests results for isolated columns, has generated a large enough database, that allowed a subsequent Bayesian updating of the methods for calculating the bearing capacity of isolated stone columns. The Bayesian updating framework has been useful to improve the predictions of the ultimate load capacity of the column, allowing to "update" the parameters of the calculation model as additional load tests become available for a specific project. Moreover, it is a valuable tool for decision making under uncertainty since it is possible to compare the cost of further testing to the cost of a possible failure and therefore to decide whether it is appropriate to perform such tests.

Monte Carlo Neutronics and Thermal-hydraulics coupling applied to Fast Reactors

Un escenario habitualmente considerado para el uso sostenible y prolongado de la energía nuclear contempla un parque de reactores rápidos refrigerados por metales líquidos (LMFR) dedicados al reciclado de Pu y la transmutación de actínidos minoritarios (MA). Otra opción es combinar dichos reactores con algunos sistemas subcríticos asistidos por acelerador (ADS), exclusivamente destinados a la eliminación de MA. El diseño y licenciamiento de estos reactores innovadores requiere herramientas computacionales prácticas y precisas, que incorporen el conocimiento obtenido en la investigación experimental de nuevas configuraciones de reactores, materiales y sistemas. A pesar de que se han construido y operado un cierto número de reactores rápidos a nivel mundial, la experiencia operacional es todavía reducida y no todos los transitorios se han podido entender completamente. Por tanto, los análisis de seguridad de nuevos LMFR están basados fundamentalmente en métodos deterministas, al contrario que las aproximaciones modernas para reactores de agua ligera (LWR), que se benefician también de los métodos probabilistas. La aproximación más usada en los estudios de seguridad de LMFR es utilizar una variedad de códigos, desarrollados a base de distintas teorías, en busca de soluciones integrales para los transitorios e incluyendo incertidumbres. En este marco, los nuevos códigos para cálculos de mejor estimación ("best estimate") que no incluyen aproximaciones conservadoras, son de una importancia primordial para analizar estacionarios y transitorios en reactores rápidos. Esta tesis se centra en el desarrollo de un código acoplado para realizar análisis realistas en reactores rápidos críticos aplicando el método de Monte Carlo. Hoy en día, dado el mayor potencial de recursos computacionales, los códigos de transporte neutrónico por Monte Carlo se pueden usar de manera práctica para realizar cálculos detallados de núcleos completos, incluso de elevada heterogeneidad material. Además, los códigos de Monte Carlo se toman normalmente como referencia para los códigos deterministas de difusión en multigrupos en aplicaciones con reactores rápidos, porque usan secciones eficaces punto a punto, un modelo geométrico exacto y tienen en cuenta intrínsecamente la dependencia angular de flujo. En esta tesis se presenta una metodología de acoplamiento entre el conocido código MCNP, que calcula la generación de potencia en el reactor, y el código de termohidráulica de subcanal COBRA-IV, que obtiene las distribuciones de temperatura y densidad en el sistema. COBRA-IV es un código apropiado para aplicaciones en reactores rápidos ya que ha sido validado con resultados experimentales en haces de barras con sodio, incluyendo las correlaciones más apropiadas para metales líquidos. En una primera fase de la tesis, ambos códigos se han acoplado en estado estacionario utilizando un método iterativo con intercambio de archivos externos. El principal problema en el acoplamiento neutrónico y termohidráulico en estacionario con códigos de Monte Carlo es la manipulación de las secciones eficaces para tener en cuenta el ensanchamiento Doppler cuando la temperatura del combustible aumenta. Entre todas las opciones disponibles, en esta tesis se ha escogido la aproximación de pseudo materiales, y se ha comprobado que proporciona resultados aceptables en su aplicación con reactores rápidos. Por otro lado, los cambios geométricos originados por grandes gradientes de temperatura en el núcleo de reactores rápidos resultan importantes para la neutrónica como consecuencia del elevado recorrido libre medio del neutrón en estos sistemas. Por tanto, se ha desarrollado un módulo adicional que simula la geometría del reactor en caliente y permite estimar la reactividad debido a la expansión del núcleo en un transitorio. éste módulo calcula automáticamente la longitud del combustible, el radio de la vaina, la separación de los elementos de combustible y el radio de la placa soporte en función de la temperatura. éste efecto es muy relevante en transitorios sin inserción de bancos de parada. También relacionado con los cambios geométricos, se ha implementado una herramienta que, automatiza el movimiento de las barras de control en busca d la criticidad del reactor, o bien calcula el valor de inserción axial las barras de control. Una segunda fase en la plataforma de cálculo que se ha desarrollado es la simulació dinámica. Puesto que MCNP sólo realiza cálculos estacionarios para sistemas críticos o supercríticos, la solución más directa que se propone sin modificar el código fuente de MCNP es usar la aproximación de factorización de flujo, que resuelve por separado la forma del flujo y la amplitud. En este caso se han estudiado en profundidad dos aproximaciones: adiabática y quasiestática. El método adiabático usa un esquema de acoplamiento que alterna en el tiempo los cálculos neutrónicos y termohidráulicos. MCNP calcula el modo fundamental de la distribución de neutrones y la reactividad al final de cada paso de tiempo, y COBRA-IV calcula las propiedades térmicas en el punto intermedio de los pasos de tiempo. La evolución de la amplitud de flujo se calcula resolviendo las ecuaciones de cinética puntual. Este método calcula la reactividad estática en cada paso de tiempo que, en general, difiere de la reactividad dinámica que se obtendría con la distribución de flujo exacta y dependiente de tiempo. No obstante, para entornos no excesivamente alejados de la criticidad ambas reactividades son similares y el método conduce a resultados prácticos aceptables. Siguiendo esta línea, se ha desarrollado después un método mejorado para intentar tener en cuenta el efecto de la fuente de neutrones retardados en la evolución de la forma del flujo durante el transitorio. El esquema consiste en realizar un cálculo cuasiestacionario por cada paso de tiempo con MCNP. La simulación cuasiestacionaria se basa EN la aproximación de fuente constante de neutrones retardados, y consiste en dar un determinado peso o importancia a cada ciclo computacial del cálculo de criticidad con MCNP para la estimación del flujo final. Ambos métodos se han verificado tomando como referencia los resultados del código de difusión COBAYA3 frente a un ejercicio común y suficientemente significativo. Finalmente, con objeto de demostrar la posibilidad de uso práctico del código, se ha simulado un transitorio en el concepto de reactor crítico en fase de diseño MYRRHA/FASTEF, de 100 MW de potencia térmica y refrigerado por plomo-bismuto. ABSTRACT Long term sustainable nuclear energy scenarios envisage a fleet of Liquid Metal Fast Reactors (LMFR) for the Pu recycling and minor actinides (MAs) transmutation or combined with some accelerator driven systems (ADS) just for MAs elimination. Design and licensing of these innovative reactor concepts require accurate computational tools, implementing the knowledge obtained in experimental research for new reactor configurations, materials and associated systems. Although a number of fast reactor systems have already been built, the operational experience is still reduced, especially for lead reactors, and not all the transients are fully understood. The safety analysis approach for LMFR is therefore based only on deterministic methods, different from modern approach for Light Water Reactors (LWR) which also benefit from probabilistic methods. Usually, the approach adopted in LMFR safety assessments is to employ a variety of codes, somewhat different for the each other, to analyze transients looking for a comprehensive solution and including uncertainties. In this frame, new best estimate simulation codes are of prime importance in order to analyze fast reactors steady state and transients. This thesis is focused on the development of a coupled code system for best estimate analysis in fast critical reactor. Currently due to the increase in the computational resources, Monte Carlo methods for neutrons transport can be used for detailed full core calculations. Furthermore, Monte Carlo codes are usually taken as reference for deterministic diffusion multigroups codes in fast reactors applications because they employ point-wise cross sections in an exact geometry model and intrinsically account for directional dependence of the ux. The coupling methodology presented here uses MCNP to calculate the power deposition within the reactor. The subchannel code COBRA-IV calculates the temperature and density distribution within the reactor. COBRA-IV is suitable for fast reactors applications because it has been validated against experimental results in sodium rod bundles. The proper correlations for liquid metal applications have been added to the thermal-hydraulics program. Both codes are coupled at steady state using an iterative method and external files exchange. The main issue in the Monte Carlo/thermal-hydraulics steady state coupling is the cross section handling to take into account Doppler broadening when temperature rises. Among every available options, the pseudo materials approach has been chosen in this thesis. This approach obtains reasonable results in fast reactor applications. Furthermore, geometrical changes caused by large temperature gradients in the core, are of major importance in fast reactor due to the large neutron mean free path. An additional module has therefore been included in order to simulate the reactor geometry in hot state or to estimate the reactivity due to core expansion in a transient. The module automatically calculates the fuel length, cladding radius, fuel assembly pitch and diagrid radius with the temperature. This effect will be crucial in some unprotected transients. Also related to geometrical changes, an automatic control rod movement feature has been implemented in order to achieve a just critical reactor or to calculate control rod worth. A step forward in the coupling platform is the dynamic simulation. Since MCNP performs only steady state calculations for critical systems, the more straight forward option without modifying MCNP source code, is to use the flux factorization approach solving separately the flux shape and amplitude. In this thesis two options have been studied to tackle time dependent neutronic simulations using a Monte Carlo code: adiabatic and quasistatic methods. The adiabatic methods uses a staggered time coupling scheme for the time advance of neutronics and the thermal-hydraulics calculations. MCNP computes the fundamental mode of the neutron flux distribution and the reactivity at the end of each time step and COBRA-IV the thermal properties at half of the the time steps. To calculate the flux amplitude evolution a solver of the point kinetics equations is used. This method calculates the static reactivity in each time step that in general is different from the dynamic reactivity calculated with the exact flux distribution. Nevertheless, for close to critical situations, both reactivities are similar and the method leads to acceptable practical results. In this line, an improved method as an attempt to take into account the effect of delayed neutron source in the transient flux shape evolutions is developed. The scheme performs a quasistationary calculation per time step with MCNP. This quasistationary simulations is based con the constant delayed source approach, taking into account the importance of each criticality cycle in the final flux estimation. Both adiabatic and quasistatic methods have been verified against the diffusion code COBAYA3, using a theoretical kinetic exercise. Finally, a transient in a critical 100 MWth lead-bismuth-eutectic reactor concept is analyzed using the adiabatic method as an application example in a real system.

Aplicación de la metodología ISA al análisis de la secuencia de pérdida del sistema de agua de refrigeración de componentes en una central nuclear con reactor de agua a presión

La metodología Integrated Safety Analysis (ISA), desarrollada en el área de Modelación y Simulación (MOSI) del Consejo de Seguridad Nuclear (CSN), es un método de Análisis Integrado de Seguridad que está siendo evaluado y analizado mediante diversas aplicaciones impulsadas por el CSN; el análisis integrado de seguridad, combina las técnicas evolucionadas de los análisis de seguridad al uso: deterministas y probabilistas. Se considera adecuado para sustentar la Regulación Informada por el Riesgo (RIR), actual enfoque dado a la seguridad nuclear y que está siendo desarrollado y aplicado en todo el mundo. En este contexto se enmarcan, los proyectos Safety Margin Action Plan (SMAP) y Safety Margin Assessment Application (SM2A), impulsados por el Comité para la Seguridad de las Instalaciones Nucleares (CSNI) de la Agencia de la Energía Nuclear (NEA) de la Organización para la Cooperación y el Desarrollo Económicos (OCDE) en el desarrollo del enfoque adecuado para el uso de las metodologías integradas en la evaluación del cambio en los márgenes de seguridad debidos a cambios en las condiciones de las centrales nucleares. El comité constituye un foro para el intercambio de información técnica y de colaboración entre las organizaciones miembro, que aportan sus propias ideas en investigación, desarrollo e ingeniería. La propuesta del CSN es la aplicación de la metodología ISA, especialmente adecuada para el análisis según el enfoque desarrollado en el proyecto SMAP que pretende obtener los valores best-estimate con incertidumbre de las variables de seguridad que son comparadas con los límites de seguridad, para obtener la frecuencia con la que éstos límites son superados. La ventaja que ofrece la ISA es que permite el análisis selectivo y discreto de los rangos de los parámetros inciertos que tienen mayor influencia en la superación de los límites de seguridad, o frecuencia de excedencia del límite, permitiendo así evaluar los cambios producidos por variaciones en el diseño u operación de la central que serían imperceptibles o complicados de cuantificar con otro tipo de metodologías. La ISA se engloba dentro de las metodologías de APS dinámico discreto que utilizan la generación de árboles de sucesos dinámicos (DET) y se basa en la Theory of Stimulated Dynamics (TSD), teoría de fiabilidad dinámica simplificada que permite la cuantificación del riesgo de cada una de las secuencias. Con la ISA se modelan y simulan todas las interacciones relevantes en una central: diseño, condiciones de operación, mantenimiento, actuaciones de los operadores, eventos estocásticos, etc. Por ello requiere la integración de códigos de: simulación termohidráulica y procedimientos de operación; delineación de árboles de sucesos; cuantificación de árboles de fallos y sucesos; tratamiento de incertidumbres e integración del riesgo. La tesis contiene la aplicación de la metodología ISA al análisis integrado del suceso iniciador de la pérdida del sistema de refrigeración de componentes (CCWS) que genera secuencias de pérdida de refrigerante del reactor a través de los sellos de las bombas principales del circuito de refrigerante del reactor (SLOCA). Se utiliza para probar el cambio en los márgenes, con respecto al límite de la máxima temperatura de pico de vaina (1477 K), que sería posible en virtud de un potencial aumento de potencia del 10 % en el reactor de agua a presión de la C.N. Zion. El trabajo realizado para la consecución de la tesis, fruto de la colaboración de la Escuela Técnica Superior de Ingenieros de Minas y Energía y la empresa de soluciones tecnológicas Ekergy Software S.L. (NFQ Solutions) con el área MOSI del CSN, ha sido la base para la contribución del CSN en el ejercicio SM2A. Este ejercicio ha sido utilizado como evaluación del desarrollo de algunas de las ideas, sugerencias, y los algoritmos detrás de la metodología ISA. Como resultado se ha obtenido un ligero aumento de la frecuencia de excedencia del daño (DEF) provocado por el aumento de potencia. Este resultado demuestra la viabilidad de la metodología ISA para obtener medidas de las variaciones en los márgenes de seguridad que han sido provocadas por modificaciones en la planta. También se ha mostrado que es especialmente adecuada en escenarios donde los eventos estocásticos o las actuaciones de recuperación o mitigación de los operadores pueden tener un papel relevante en el riesgo. Los resultados obtenidos no tienen validez más allá de la de mostrar la viabilidad de la metodología ISA. La central nuclear en la que se aplica el estudio está clausurada y la información relativa a sus análisis de seguridad es deficiente, por lo que han sido necesarias asunciones sin comprobación o aproximaciones basadas en estudios genéricos o de otras plantas. Se han establecido tres fases en el proceso de análisis: primero, obtención del árbol de sucesos dinámico de referencia; segundo, análisis de incertidumbres y obtención de los dominios de daño; y tercero, cuantificación del riesgo. Se han mostrado diversas aplicaciones de la metodología y ventajas que presenta frente al APS clásico. También se ha contribuido al desarrollo del prototipo de herramienta para la aplicación de la metodología ISA (SCAIS). ABSTRACT The Integrated Safety Analysis methodology (ISA), developed by the Consejo de Seguridad Nuclear (CSN), is being assessed in various applications encouraged by CSN. An Integrated Safety Analysis merges the evolved techniques of the usually applied safety analysis methodologies; deterministic and probabilistic. It is considered as a suitable tool for assessing risk in a Risk Informed Regulation framework, the approach under development that is being adopted on Nuclear Safety around the world. In this policy framework, the projects Safety Margin Action Plan (SMAP) and Safety Margin Assessment Application (SM2A), set up by the Committee on the Safety of Nuclear Installations (CSNI) of the Nuclear Energy Agency within the Organization for Economic Co-operation and Development (OECD), were aimed to obtain a methodology and its application for the integration of risk and safety margins in the assessment of the changes to the overall safety as a result of changes in the nuclear plant condition. The committee provides a forum for the exchange of technical information and cooperation among member organizations which contribute their respective approaches in research, development and engineering. The ISA methodology, proposed by CSN, specially fits with the SMAP approach that aims at obtaining Best Estimate Plus Uncertainty values of the safety variables to be compared with the safety limits. This makes it possible to obtain the exceedance frequencies of the safety limit. The ISA has the advantage over other methods of allowing the specific and discrete evaluation of the most influential uncertain parameters in the limit exceedance frequency. In this way the changes due to design or operation variation, imperceptibles or complicated to by quantified by other methods, are correctly evaluated. The ISA methodology is one of the discrete methodologies of the Dynamic PSA framework that uses the generation of dynamic event trees (DET). It is based on the Theory of Stimulated Dynamics (TSD), a simplified version of the theory of Probabilistic Dynamics that allows the risk quantification. The ISA models and simulates all the important interactions in a Nuclear Power Plant; design, operating conditions, maintenance, human actuations, stochastic events, etc. In order to that, it requires the integration of codes to obtain: Thermohydraulic and human actuations; Even trees delineation; Fault Trees and Event Trees quantification; Uncertainty analysis and risk assessment. This written dissertation narrates the application of the ISA methodology to the initiating event of the Loss of the Component Cooling System (CCWS) generating sequences of loss of reactor coolant through the seals of the reactor coolant pump (SLOCA). It is used to test the change in margins with respect to the maximum clad temperature limit (1477 K) that would be possible under a potential 10 % power up-rate effected in the pressurized water reactor of Zion NPP. The work done to achieve the thesis, fruit of the collaborative agreement of the School of Mining and Energy Engineering and the company of technological solutions Ekergy Software S.L. (NFQ Solutions) with de specialized modeling and simulation branch of the CSN, has been the basis for the contribution of the CSN in the exercise SM2A. This exercise has been used as an assessment of the development of some of the ideas, suggestions, and algorithms behind the ISA methodology. It has been obtained a slight increase in the Damage Exceedance Frequency (DEF) caused by the power up-rate. This result shows that ISA methodology allows quantifying the safety margin change when design modifications are performed in a NPP and is specially suitable for scenarios where stochastic events or human responses have an important role to prevent or mitigate the accidental consequences and the total risk. The results do not have any validity out of showing the viability of the methodology ISA. Zion NPP was retired and information of its safety analysis is scarce, so assumptions without verification or approximations based on generic studies have been required. Three phases are established in the analysis process: first, obtaining the reference dynamic event tree; second, uncertainty analysis and obtaining the damage domains; third, risk quantification. There have been shown various applications of the methodology and advantages over the classical PSA. It has also contributed to the development of the prototype tool for the implementation of the ISA methodology (SCAIS).

Análisis de secuencias de accidente de pérdida de refrigerante en centrales nucleares de agua a presión mediante la metodología ISA

El accidente de pérdida de refrigerante (LOCA) en un reactor nuclear es uno de los accidentes Base de Diseño más preocupantes y estudiados desde el origen del uso de la tecnología de fisión en la industria productora de energía. El LOCA ocupa, desde el punto de vista de los análisis de seguridad, un lugar de vanguardia tanto en el análisis determinista (DSA) como probabilista (PSA), cuya diferenciada perspectiva ha ido evolucionando notablemente en lo que al crédito a la actuación de las salvaguardias y las acciones del operador se refiere. En la presente tesis se aborda el análisis sistemático de de las secuencias de LOCA por pequeña y mediana rotura en diferentes lugares de un reactor nuclear de agua a presión (PWR) con fallo total de Inyección de Seguridad de Alta Presión (HPSI). Tal análisis ha sido desarrollado en base a la metodología de Análisis Integrado de Seguridad (ISA), desarrollado por el Consejo de Seguridad Nuclear (CSN) y consistente en la aplicación de métodos avanzados de simulación y PSA para la obtención de Dominios de Daño, que cuantifican topológicamente las probabilidades de éxito y daño en función de determinados parámetros inciertos. Para la elaboración de la presente tesis, se ha hecho uso del código termohidráulico TRACE v5.0 (patch 2), avalado por la NRC de los EEUU como código de planta para la simulación y análisis de secuencias en reactores de agua ligera (LWR). Los objetivos del trabajo son, principalmente: (1) el análisis exhaustivo de las secuencias de LOCA por pequeña-mediana rotura en diferentes lugares de un PWR de tres lazos de diseño Westinghouse (CN Almaraz), con fallo de HPSI, en función de parámetros de gran importancia para los transitorios, tales como el tamaño de rotura y el tiempo de retraso en la respuesta del operador; (2) la obtención y análisis de los Dominios de Daño para transitorios de LOCA en PWRs, de acuerdo con la metodología ISA; y (3) la revisión de algunos de los resultados genéricos de los análisis de seguridad para secuencias de LOCA en las mencionadas condiciones. Los resultados de la tesis abarcan tres áreas bien diferenciadas a lo largo del trabajo: (a) la fenomenología física de las secuencias objeto de estudio; (b) las conclusiones de los análisis de seguridad practicados a los transitorios de LOCA; y (c) la relevancia de las consecuencias de las acciones humanas por parte del grupo de operación. Estos resultados, a su vez, son de dos tipos fundamentales: (1) de respaldo del conocimiento previo sobre el tipo de secuencias analizado, incluido en la extensa bibliografía examinada; y (2) hallazgos en cada una de las tres áreas mencionadas, no referidos en la bibliografía. En resumidas cuentas, los resultados de la tesis avalan el uso de la metodología ISA como método de análisis alternativo y sistemático para secuencias accidentales en LWRs. ABSTRACT The loss of coolant accident (LOCA) in nuclear reactors is one of the most concerning and analized accidents from the beginning of the use of fission technology for electric power production. From the point of view of safety analyses, LOCA holds a forefront place in both Deterministic (DSA) and Probabilistic Safety Analysis (PSA), which have significantly evolved from their original state in both safeguard performance credibility and human actuation. This thesis addresses a systematic analysis of small and medium LOCA sequences, in different places of a nuclear Pressurized Water Reactor (PWR) and with total failure of High Pressure Safety Injection (HPSI). Such an analysis has been grounded on the Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Regulatory Body (CSN). ISA involves the application of advanced methods of simulation and PSA for obtaining Damage Domains that topologically quantify the likelihood of success and damage regarding certain uncertain parameters.TRACE v5.0 (patch 2) code has been used as the thermalhydraulic simulation tool for the elaboration of this work. Nowadays, TRACE is supported by the US NRC as a plant code for the simulation and analysis of sequences in light water reactors (LWR). The main objectives of the work are the following ones: (1) the in-depth analysis of small and medium LOCA sequences in different places of a Westinghouse three-loop PWR (Almaraz NPP), with failed HPSI, regarding important parameters, such as break size or delay in operator response; (2) obtainment and analysis of Damage Domains related to LOCA transients in PWRs, according to ISA methodology; and (3) review some of the results of generic safety analyses for LOCA sequences in those conditions. The results of the thesis cover three separated areas: (a) the physical phenomenology of the sequences under study; (b) the conclusions of LOCA safety analyses; and (c) the importance of consequences of human actions by the operating crew. These results, in turn, are of two main types: (1) endorsement of previous knowledge about this kind of sequences, which is included in the literature; and (2) findings in each of the three aforementioned areas, not reported in the reviewed literature. In short, the results of this thesis support the use of ISA-like methodology as an alternative method for systematic analysis of LWR accidental sequences.

Model-Driven safety evaluation with state-event-based component failure annotations

Over the past years, the paradigm of component-based software engineering has been established in the construction of complex mission-critical systems. Due to this trend, there is a practical need for techniques that evaluate critical properties (such as safety, reliability, availability or performance) of these systems. In this paper, we review several high-level techniques for the evaluation of safety properties for component-based systems and we propose a new evaluation model (State Event Fault Trees) that extends safety analysis towards a lower abstraction level. This model possesses a state-event semantics and strong encapsulation, which is especially useful for the evaluation of component-based software systems. Finally, we compare the techniques and give suggestions for their combined usage

An automated failure mode and effect analysis based on high-level design specificication with behavior trees

Formal methods have significant benefits for developing safety critical systems, in that they allow for correctness proofs, model checking safety and liveness properties, deadlock checking, etc. However, formal methods do not scale very well and demand specialist skills, when developing real-world systems. For these reasons, development and analysis of large-scale safety critical systems will require effective integration of formal and informal methods. In this paper, we use such an integrative approach to automate Failure Modes and Effects Analysis (FMEA), a widely used system safety analysis technique, using a high-level graphical modelling notation (Behavior Trees) and model checking. We inject component failure modes into the Behavior Trees and translate the resulting Behavior Trees to SAL code. This enables us to model check if the system in the presence of these faults satisfies its safety properties, specified by temporal logic formulas. The benefit of this process is tool support that automates the tedious and error-prone aspects of FMEA.

Development of safety performance functions for SafetyAnalyst applications in Florida

In 2010, the American Association of State Highway and Transportation Officials (AASHTO) released a safety analysis software system known as SafetyAnalyst. SafetyAnalyst implements the empirical Bayes (EB) method, which requires the use of Safety Performance Functions (SPFs). The system is equipped with a set of national default SPFs, and the software calibrates the default SPFs to represent the agency's safety performance. However, it is recommended that agencies generate agency-specific SPFs whenever possible. Many investigators support the view that the agency-specific SPFs represent the agency data better than the national default SPFs calibrated to agency data. Furthermore, it is believed that the crash trends in Florida are different from the states whose data were used to develop the national default SPFs. In this dissertation, Florida-specific SPFs were developed using the 2008 Roadway Characteristics Inventory (RCI) data and crash and traffic data from 2007-2010 for both total and fatal and injury (FI) crashes. The data were randomly divided into two sets, one for calibration (70% of the data) and another for validation (30% of the data). The negative binomial (NB) model was used to develop the Florida-specific SPFs for each of the subtypes of roadway segments, intersections and ramps, using the calibration data. Statistical goodness-of-fit tests were performed on the calibrated models, which were then validated using the validation data set. The results were compared in order to assess the transferability of the Florida-specific SPF models. The default SafetyAnalyst SPFs were calibrated to Florida data by adjusting the national default SPFs with local calibration factors. The performance of the Florida-specific SPFs and SafetyAnalyst default SPFs calibrated to Florida data were then compared using a number of methods, including visual plots and statistical goodness-of-fit tests. The plots of SPFs against the observed crash data were used to compare the prediction performance of the two models. Three goodness-of-fit tests, represented by the mean absolute deviance (MAD), the mean square prediction error (MSPE), and Freeman-Tukey R2 (R2FT), were also used for comparison in order to identify the better-fitting model. The results showed that Florida-specific SPFs yielded better prediction performance than the national default SPFs calibrated to Florida data. The performance of Florida-specific SPFs was further compared with that of the full SPFs, which include both traffic and geometric variables, in two major applications of SPFs, i.e., crash prediction and identification of high crash locations. The results showed that both SPF models yielded very similar performance in both applications. These empirical results support the use of the flow-only SPF models adopted in SafetyAnalyst, which require much less effort to develop compared to full SPFs.

Caracterização de crime ambiental de poluição por meio de abordagem multiparamétrica e incorporando incerteza de amostragem

Tese (Doutorado em Tecnologia Nuclear)

Development of Safety Performance Functions for SafetyAnalyst Applications in Florida

In 2010, the American Association of State Highway and Transportation Officials (AASHTO) released a safety analysis software system known as SafetyAnalyst. SafetyAnalyst implements the empirical Bayes (EB) method, which requires the use of Safety Performance Functions (SPFs). The system is equipped with a set of national default SPFs, and the software calibrates the default SPFs to represent the agency’s safety performance. However, it is recommended that agencies generate agency-specific SPFs whenever possible. Many investigators support the view that the agency-specific SPFs represent the agency data better than the national default SPFs calibrated to agency data. Furthermore, it is believed that the crash trends in Florida are different from the states whose data were used to develop the national default SPFs. In this dissertation, Florida-specific SPFs were developed using the 2008 Roadway Characteristics Inventory (RCI) data and crash and traffic data from 2007-2010 for both total and fatal and injury (FI) crashes. The data were randomly divided into two sets, one for calibration (70% of the data) and another for validation (30% of the data). The negative binomial (NB) model was used to develop the Florida-specific SPFs for each of the subtypes of roadway segments, intersections and ramps, using the calibration data. Statistical goodness-of-fit tests were performed on the calibrated models, which were then validated using the validation data set. The results were compared in order to assess the transferability of the Florida-specific SPF models. The default SafetyAnalyst SPFs were calibrated to Florida data by adjusting the national default SPFs with local calibration factors. The performance of the Florida-specific SPFs and SafetyAnalyst default SPFs calibrated to Florida data were then compared using a number of methods, including visual plots and statistical goodness-of-fit tests. The plots of SPFs against the observed crash data were used to compare the prediction performance of the two models. Three goodness-of-fit tests, represented by the mean absolute deviance (MAD), the mean square prediction error (MSPE), and Freeman-Tukey R2 (R2FT), were also used for comparison in order to identify the better-fitting model. The results showed that Florida-specific SPFs yielded better prediction performance than the national default SPFs calibrated to Florida data. The performance of Florida-specific SPFs was further compared with that of the full SPFs, which include both traffic and geometric variables, in two major applications of SPFs, i.e., crash prediction and identification of high crash locations. The results showed that both SPF models yielded very similar performance in both applications. These empirical results support the use of the flow-only SPF models adopted in SafetyAnalyst, which require much less effort to develop compared to full SPFs.

Avoiding Divergence in the Shalvi-Weinstein Algorithm

The most popular algorithms for blind equalization are the constant-modulus algorithm (CMA) and the Shalvi-Weinstein algorithm (SWA). It is well-known that SWA presents a higher convergence rate than CMA. at the expense of higher computational complexity. If the forgetting factor is not sufficiently close to one, if the initialization is distant from the optimal solution, or if the signal-to-noise ratio is low, SWA can converge to undesirable local minima or even diverge. In this paper, we show that divergence can be caused by an inconsistency in the nonlinear estimate of the transmitted signal. or (when the algorithm is implemented in finite precision) by the loss of positiveness of the estimate of the autocorrelation matrix, or by a combination of both. In order to avoid the first cause of divergence, we propose a dual-mode SWA. In the first mode of operation. the new algorithm works as SWA; in the second mode, it rejects inconsistent estimates of the transmitted signal. Assuming the persistence of excitation condition, we present a deterministic stability analysis of the new algorithm. To avoid the second cause of divergence, we propose a dual-mode lattice SWA, which is stable even in finite-precision arithmetic, and has a computational complexity that increases linearly with the number of adjustable equalizer coefficients. The good performance of the proposed algorithms is confirmed through numerical simulations.

A Sustained Virologic Response Is Durable in Patients With Chronic Hepatitis C Treated With Peginterferon Alfa-2a and Ribavirin

BACKGROUND & AIMS: A sustained virologic response (SVR) to therapy for hepatitis C virus (HCV) infection is defined as the inability to detect HCV RNA 24 weeks after completion of treatment. Although small studies have reported that the SVR is durable and lasts for long periods, it has not been conclusively shown. METHODS: The durability of treatment responses was examined in patients originally enrolled in one of 9 randomized multicenter trials (n = 1343). The study included patients who received pegylated interferon (peginterferon) alfa-2a alone (n = 166) or in combination with ribavirin (n = 1077, including 79 patients with normal alanine aminotransferase levels and 100 patients who were coinfected with human immunodeficiency virus and HCV) and whose serum samples were negative for HCV RNA (<50 IU/mL) at their final assessment. Patients were assessed annually, from the date of last treatment, for a mean of 3.9 years (range, 0.8-7.1 years). RESULTS: Most patients (99.1%) who achieved an SVR had undetectable levels of HCV RNA in serum samples throughout the follow-up period. Serum samples from 0.9% of the patients contained HCV RNA a mean of 1.8 years (range, 1.1-2.9 years) after treatment ended. It is not clear if these patients were reinfected or experienced a relapse. CONCLUSIONS: In a large cohort of patients monitored for the durability of an SVR, the SVR was maintained for almost 4 years after treatment with peginterferon alfa-2a alone or in combination with ribavirin. In patients with chronic hepatitis C infection, the SVR is durable and these patients should be considered as cured.

Análise probabilística de segurança de barragens gravidade de betão

Trabalho Final de Mestrado para obtenção do grau de Mestre em Engenharia Civil na Área de Especialização de Estruturas

Intervenção para a segurança do doente: análise dos obstáculos existentes na transferência de informação durante a passagem de turno de enfermagem

Mestrado em Intervenção Sócio-Organizacional na Saúde - Ramo de especialização: Políticas de Administração e Gestão de Serviços de Saúde

Probabilistic maintenance and optimization strategies for deteriorating civil infrastructures

In developed countries, civil infrastructures are one of the most significant investments of governments, corporations, and individuals. Among these, transportation infrastructures, including highways, bridges, airports, and ports, are of huge importance, both economical and social. Most developed countries have built a fairly complete network of highways to fit their needs. As a result, the required investment in building new highways has diminished during the last decade, and should be further reduced in the following years. On the other hand, significant structural deteriorations have been detected in transportation networks, and a huge investment is necessary to keep these infrastructures safe and serviceable. Due to the significant importance of bridges in the serviceability of highway networks, maintenance of these structures plays a major role. In this paper, recent progress in probabilistic maintenance and optimization strategies for deteriorating civil infrastructures with emphasis on bridges is summarized. A novel model including interaction between structural safety analysis,through the safety index, and visual inspections and non destructive tests, through the condition index, is presented. Single objective optimization techniques leading to maintenance strategies associated with minimum expected cumulative cost and acceptable levels of condition and safety are presented. Furthermore, multi-objective optimization is used to simultaneously consider several performance indicators such as safety, condition, and cumulative cost. Realistic examples of the application of some of these techniques and strategies are also presented.

Análise e avaliação de riscos ocupacionais em actividades de montagem eléctrica

Dissertação para obtenção do Grau de Mestre em Engenharia e Gestão Industrial