Education and Training of Future Nuclear Engineers Through the use of an Interactive Plant Simulator

The successful experience of the Jose Cabrera Nuclear Power Plant Interactive Graphical Simulator implementation in the Nuclear Engineering Department in the Universidad Polite´cnica de Madrid, for the Education and Training of nuclear engineers is shown in this paper. The paper starts with the objectives and the description of the Simulator Aula, and the methodology of work following the recommendations of the IAEA for the use of nuclear reactor simulators for education. The practices and material prepared for the students, as well as the operational and accident situations simulated are provided.

Interactive Graphic Simulation: An Advanced Methodology to Improve the Teaching-Learning Process in Nuclear Engineering Education and Training

Nowadays, computer simulators are becoming basic tools for education and training in many engineering fields. In the nuclear industry, the role of simulation for training of operators of nuclear power plants is also recognized of the utmost relevance. As an example, the International Atomic Energy Agency sponsors the development of nuclear reactor simulators for education, and arranges the supply of such simulation programs. Aware of this, in 2008 Gas Natural Fenosa, a Spanish gas and electric utility that owns and operate nuclear power plants and promotes university education in the nuclear technology field, provided the Department of Nuclear Engineering of Universidad Politécnica de Madrid with the Interactive Graphic Simulator (IGS) of “José Cabrera” (Zorita) nuclear power plant, an industrial facility whose commercial operation ceased definitively in April 2006. It is a state-of-the-art full-scope real-time simulator that was used for training and qualification of the operators of the plant control room, as well as to understand and analyses the plant dynamics, and to develop, qualify and validate its emergency operating procedures.

Pond head level control in a run-of-river hydro power plant using fuzzy controller

The run-of-river hydro power plant usually have low or nil water storage capacity, and therefore an adequate control strategy is required to keep the water level constant in pond. This paper presents a novel technique based on TSK fuzzy controller to maintain the pond head constant. The performance is investigated over a wide range of hill curve of hydro turbine. The results are compared with PI controller as discussed in [1].

Real time power plant simulation platform for training on Electrical Protections and Automatic Voltage Regulators

Electrical Protection systems and Automatic Voltage Regulators (AVR) are essential components of actual power plants. Its installation and setting is performed during the commissioning, and it needs extensive experience since any failure in this process or in the setting, may entails some risk not only for the generator of the power plant, but also for the reliability of the power grid. In this paper, a real time power plant simulation platform is presented as a tool for improving the training and learning process on electrical protections and automatic voltage regulators. The activities of the commissioning procedure which can be practiced are described, and the applicability of this tool for improving the comprehension of this important part of the power plants is discussed. A commercial AVR and a multifunction protective relay have been tested with satisfactory results.

Distribution of the coal flow in the mill-duct system of the As Pontes Power Plant using CFD modeling

The efficiency of a Power Plant is affected by the distribution of the pulverized coal within the furnace. The coal, which is pulverized in the mills, is transported and distributed by the primary gas through the mill-ducts to the interior of the furnace. This is done with a double function: dry and enter the coal by different levels for optimizing the combustion in the sense that a complete combustion occurs with homogeneous heat fluxes to the walls. The mill-duct systems of a real Power Plant are very complex and they are not yet well understood. In particular, experimental data concerning the mass flows of coal to the different levels are very difficult to measure. CFD modeling can help to determine them. An Eulerian/Lagrangian approach is used due to the low solid–gas volume ratio.

Conceptual design of offshore platform supply vessel based on hybrid diesel generator-fuel cell power plant

Nowadays increasing fuel prices and upcoming pollutant emission regulations are becoming a growing concern for the shipping industry worldwide. While fuel prices will keep rising in future years, the new International Convention for the Prevention of Pollution from Ships (MARPOL) and Sulphur Emissions Control Areas (SECA) regulations will forbid ships to use heavy fuel oils at certain situations. To fulfil with these regulations, the next step in the marine shipping business will comprise the use of cleaner fuels on board as well as developing new propulsion concept. In this work a new conceptual marine propulsion system is developed, based on the integration of diesel generators with fuel cells in a 2850 metric tonne of deadweight platform supply vessel. The efficiency of the two 250 kW methanol-fed Solid Oxide Fuel Cell (SOFC) system installed on board combined with the hydro dynamically optimized design of the hull of the ship will allow the ship to successfully operate at certain modes of operation while notably reduce the pollutant emissions to the atmosphere. Besides the cogeneration heat obtained from the fuel cell system will be used to answer different heating needs on board the vessel

Validation Study of the Use of Matlab/Simulink Synchronous-Machine Block for Accurate Power-Plant Stability Studies

This paper presents results of the validity study of the use of MATLAB/Simulink synchronous-machine block for power-system stability studies. Firstly, the waveforms of the theoretical synchronous-generator short-circuit currents are described. Thereafter, the comparison between the currents obtained through the simulation model in the sudden short-circuit test, are compared to the theoretical ones. Finally, the factory tests of two commercial generating units are compared to the response of the synchronous generator simulation block during sudden short-circuit, set with the same real data, with satisfactory results. This results show the validity of the use of this generator block for power plant simulation.

Hybrid wind-diesel power plant valuation and optimization using real option theory and dynamic programming

El sistema de energía eólica-diesel híbrido tiene un gran potencial en la prestación de suministro de energía a comunidades remotas. En comparación con los sistemas tradicionales de diesel, las plantas de energía híbridas ofrecen grandes ventajas tales como el suministro de capacidad de energía extra para "microgrids", reducción de los contaminantes y emisiones de gases de efecto invernadero, y la cobertura del riesgo de aumento inesperado del precio del combustible. El principal objetivo de la presente tesis es proporcionar nuevos conocimientos para la evaluación y optimización de los sistemas de energía híbrido eólico-diesel considerando las incertidumbres. Dado que la energía eólica es una variable estocástica, ésta no puede ser controlada ni predecirse con exactitud. La naturaleza incierta del viento como fuente de energía produce serios problemas tanto para la operación como para la evaluación del valor del sistema de energía eólica-diesel híbrido. Por un lado, la regulación de la potencia inyectada desde las turbinas de viento es una difícil tarea cuando opera el sistema híbrido. Por otro lado, el bene.cio económico de un sistema eólico-diesel híbrido se logra directamente a través de la energía entregada a la red de alimentación de la energía eólica. Consecuentemente, la incertidumbre de los recursos eólicos incrementa la dificultad de estimar los beneficios globales en la etapa de planificación. La principal preocupación del modelo tradicional determinista es no tener en cuenta la incertidumbre futura a la hora de tomar la decisión de operación. Con lo cual, no se prevé las acciones operativas flexibles en respuesta a los escenarios futuros. El análisis del rendimiento y simulación por ordenador en el Proyecto Eólico San Cristóbal demuestra que la incertidumbre sobre la energía eólica, las estrategias de control, almacenamiento de energía, y la curva de potencia de aerogeneradores tienen un impacto significativo sobre el rendimiento del sistema. En la presente tesis, se analiza la relación entre la teoría de valoración de opciones y el proceso de toma de decisiones. La opción real se desarrolla con un modelo y se presenta a través de ejemplos prácticos para evaluar el valor de los sistemas de energía eólica-diesel híbridos. Los resultados muestran que las opciones operacionales pueden aportar un valor adicional para el sistema de energía híbrida, cuando esta flexibilidad operativa se utiliza correctamente. Este marco se puede aplicar en la optimización de la operación a corto plazo teniendo en cuenta la naturaleza dependiente de la trayectoria de la política óptima de despacho, dadas las plausibles futuras realizaciones de la producción de energía eólica. En comparación con los métodos de valoración y optimización existentes, el resultado del caso de estudio numérico muestra que la política de operación resultante del modelo de optimización propuesto presenta una notable actuación en la reducción del con- sumo total de combustible del sistema eólico-diesel. Con el .n de tomar decisiones óptimas, los operadores de plantas de energía y los gestores de éstas no deben centrarse sólo en el resultado directo de cada acción operativa, tampoco deberían tomar decisiones deterministas. La forma correcta es gestionar dinámicamente el sistema de energía teniendo en cuenta el valor futuro condicionado en cada opción frente a la incertidumbre. ABSTRACT Hybrid wind-diesel power systems have a great potential in providing energy supply to remote communities. Compared with the traditional diesel systems, hybrid power plants are providing many advantages such as providing extra energy capacity to the micro-grid, reducing pollution and greenhouse-gas emissions, and hedging the risk of unexpected fuel price increases. This dissertation aims at providing novel insights for assessing and optimizing hybrid wind-diesel power systems considering the related uncertainties. Since wind power can neither be controlled nor accurately predicted, the energy harvested from a wind turbine may be considered a stochastic variable. This uncertain nature of wind energy source results in serious problems for both the operation and value assessment of the hybrid wind-diesel power system. On the one hand, regulating the uncertain power injected from wind turbines is a difficult task when operating the hybrid system. On the other hand, the economic profit of a hybrid wind-diesel system is achieved directly through the energy delivered to the power grid from the wind energy. Therefore, the uncertainty of wind resources has increased the difficulty in estimating the total benefits in the planning stage. The main concern of the traditional deterministic model is that it does not consider the future uncertainty when making the dispatch decision. Thus, it does not provide flexible operational actions in response to the uncertain future scenarios. Performance analysis and computer simulation on the San Cristobal Wind Project demonstrate that the wind power uncertainty, control strategies, energy storage, and the wind turbine power curve have a significant impact on the performance of the system. In this dissertation, the relationship between option pricing theory and decision making process is discussed. A real option model is developed and presented through practical examples for assessing the value of hybrid wind-diesel power systems. Results show that operational options can provide additional value to the hybrid power system when this operational flexibility is correctly utilized. This framework can be applied in optimizing short term dispatch decisions considering the path-dependent nature of the optimal dispatch policy, given the plausible future realizations of the wind power production. Comparing with the existing valuation and optimization methods, result from numerical example shows that the dispatch policy resulting from the proposed optimization model exhibits a remarkable performance in minimizing the total fuel consumption of the wind-diesel system. In order to make optimal decisions, power plant operators and managers should not just focus on the direct outcome of each operational action; neither should they make deterministic decisions. The correct way is to dynamically manage the power system by taking into consideration the conditional future value in each option in response to the uncertainty.

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).

Accident Management Actions In An Upper-Head Small-Break Loss-Of-Coolant Accident With High-Pressure Safety Injection Failed

Since the Three Mile Island accident, an important focus of pressurized water reactor (PWR) transient analyses has been a small-break loss-of-coolant accident (SBLOCA). In 2002, the discovery of thinning of the vessel head wall at the Davis Besse nuclear power plant reactor indicated the possibility of an SBLOCA in the upper head of the reactor vessel as a result of circumferential cracking of a control rod drive mechanism penetration nozzle - which has cast even greater importance on the study of SBLOCAs. Several experimental tests have been performed at the Large Scale Test Facility to simulate the behavior of a PWR during an upper-head SBLOCA. The last of these tests, Organisation for Economic Co-operation and Development Nuclear Energy Agency Rig of Safety Assessment (OECD/NEA ROSA) Test 6.1, was performed in 2005. This test was simulated with the TRACE 5.0 code, and good agreement with the experimental results was obtained. Additionally, a broad analysis of an upper-head SBLOCA with high-pressure safety injection failed in a Westinghouse PWR was performed taking into account different accident management actions and conditions in order to check their suitability. This issue has been analyzed also in the framework of the OECD/NEA ROSA project and the Code Applications and Maintenance Program (CAMP). The main conclusion is that the current emergency operating procedures for Westinghouse reactor design are adequate for these kinds of sequences, and they do not need to be modified.

Problemas de interacción suelo estructura en cimentaciones y muros de contención. Influencia de los fenómenos de despegue y deslizamiento

Los terremotos constituyen una de las más importantes fuentes productoras de cargas dinámicas que actúan sobre las estructuras y sus cimentaciones. Cuando se produce un terremoto la energía liberada genera movimientos del terreno en forma de ondas sísmicas que pueden provocar asientos en las cimentaciones de los edificios, empujes sobre los muros de contención, vuelco de las estructuras y el suelo puede licuar perdiendo su capacidad de soporte. Los efectos de los terremotos en estructuras constituyen unos de los aspectos que involucran por su condición de interacción sueloestructura, disciplinas diversas como el Análisis Estructural, la Mecánica de Suelo y la Ingeniería Sísmica. Uno de los aspectos que han sido poco estudiados en el cálculo de estructuras sometidas a la acciones de los terremotos son los efectos del comportamiento no lineal del suelo y de los movimientos que pueden producirse bajo la acción de cargas sísmicas, tales como posibles despegues y deslizamientos. En esta Tesis se estudian primero los empujes sísmicos y posibles deslizamientos de muros de contención y se comparan las predicciones de distintos tipos de cálculos: métodos pseudo-estáticos como el de Mononobe-Okabe (1929) con la contribución de Whitman-Liao (1985), y formulaciones analíticas como la desarrollada por Veletsos y Younan (1994). En segundo lugar se estudia el efecto del comportamiento no lineal del terreno en las rigideces de una losa de cimentación superficial y circular, como la correspondiente a la chimenea de una Central Térmica o al edificio del reactor de una Central Nuclear, considerando su variación con frecuencia y con el nivel de cargas. Finalmente se estudian los posibles deslizamientos y separación de las losas de estas dos estructuras bajo la acción de terremotos, siguiendo la formulación propuesta por Wolf (1988). Para estos estudios se han desarrollado una serie de programas específicos (MUROSIS, VELETSOS, INTESES y SEPARSE) cuyos listados y detalles se incluyen en los Apéndices. En el capítulo 6 se incluyen las conclusiones resultantes de estos estudios y recomendaciones para futuras investigaciones. ABSTRACT Earthquakes constitute one of the most important sources of dynamic loads that acting on structures and foundations. When an earthquake occurs the liberated energy generates seismic waves that can give rise to structural vibrations, settlements of the foundations of buildings, pressures on retaining walls, and possible sliding, uplifting or even overturning of structures. The soil can also liquefy losing its capacity of support The study of the effects of earthquakes on structures involve the use of diverse disciplines such as Structural Analysis, Soil Mechanics and Earthquake Engineering. Some aspects that have been the subject of limited research in relation to the behavior of structures subjected to earthquakes are the effects of nonlinear soil behavior and geometric nonlinearities such as sliding and uplifting of foundations. This Thesis starts with the study of the seismic pressures and potential displacements of retaining walls comparing the predictions of two types of formulations and assessing their range of applicability and limitations: pseudo-static methods as proposed by Mononobe-Okabe (1929), with the contribution of Whitman-Liao (1985), and analytical formulations as the one developed by Veletsos and Younan (1994) for rigid walls. The Thesis deals next with the effects of nonlinear soil behavior on the dynamic stiffness of circular mat foundations like the chimney of a Thermal Power Station or the reactor building of a Nuclear Power Plant, as a function of frequency and level of forces. Finally the seismic response of these two structures accounting for the potential sliding and uplifting of the foundation under a given earthquake are studied, following an approach suggested by Wolf (1988). In order to carry out these studies a number of special purposes computer programs were developed (MUROSIS, VELETSOS, INTESES and SEPARSE). The listing and details of these programs are included in the appendices. The conclusions derived from these studies and recommendations for future work are presented in Chapter 6.

Control rod drop surveillance using two friction coefficients

In the case of large burnup, a control rod (CR) guide tube in the pressurized water reactor of a commercial nuclear power plant might bend. As a consequence, a CR drop experiment may indicate an event of a CR partially inserted and whether the CR should be deemed inoperable. Early prevention of such an event can be achieved by measuring two friction coefficients: the hydraulic coefficient and the sliding coefficient. The hydraulic coefficient hardly changes, so that the curvature of the guide tube can only be detected thanks to a variation of the sliding coefficient. A simple model for the CR drop is established and validated with CR drop experiments. If tmx denotes the instant of CR maximum velocity, a linear relationship between (tmx)_2 and the sliding coefficient is found.

Impacto y protección radiológica asociados al transporte de residuos de alta actividad (RAA) al almacén temporal centralizado (ATC) en España

Las cuestiones relacionadas con el transporte de residuos radiactivos de alta actividad (RAA) al previsto almacén temporal centralizado (ATC) en Villar de Cañas (Cuenca) están de actualidad, debido a la movilidad que se espera en un futuro próximo, el compromiso con el medio ambiente, la protección de las personas, así, como la normativa legal reguladora. En esta tesis se ha evaluado el impacto radiológico asociado a este tipo de transportes mediante una nueva herramienta de procesamiento de datos, que puede ser de utilidad y servir como documentación complementaria a la recogida en el marco legal del transporte. Además puede facilitar el análisis desde una perspectiva más científica, para investigadores, responsables públicos y técnicos en general, que pueden utilizar dicha herramienta para simular distintos escenarios de transportes radiactivos basados únicamente en datos de los materiales de entrada y las rutas elegidas. Así, conociendo el nivel de radiación a un metro del transporte y eligiendo una ruta, obtendremos los impactos asociados, tales como las poblaciones afectadas, la dosis recibida por la persona más expuesta, el impacto radiológico global, las dosis a la población en el trayecto y el posible detrimento de su salud. En España se prevé una larga “ruta radiactiva” de más de 2.000 kilómetros, por la que el combustible nuclear gastado se transportará presumiblemente por carretera desde las centrales nucleares hasta el ATC, así como los residuos vitrificados procedentes del reprocesado del combustible de la central nuclear Vandellos I, que en la actualidad están en Francia. Como conclusión más importante, se observa que la emisión de radiaciones ionizantes procedentes del transporte de residuos radiactivos de alta actividad en España, en operación normal, no es significativa a la hora de generar efectos adversos en la salud humana y su impacto radiológico puede considerarse despreciable. En caso de accidente, aunque la posibilidad del suceso es remota, las emisiones, no serán determinantes a la hora de generar efectos adversos en la salud humana. Issues related to the transport of high level radioactive wastes (HLW) to the new centralised temporary storage facility to be built in Villar de Cañas (Cuenca) are attracting renewed attention due to the mobility expected in the near future for these materials, the commitment to the environment, the protection of persons and the legal regulatory standards. This study assesses the radiological impacts associated with this type of transport by means of a new dataprocessing tool, which may be of use and serve as documentation complementary to that included in the legal framework covering transport. Furthermore, it may facilitate analysis from a more scientific perspective for researchers, public servants and technicians in general, who may use the tool to simulate different radioactive transport scenarios based only on input materials data and the routes selected. Thus, by knowing the radiation level at a distance of one metre from the transport and selecting a route, it is possible to obtain the associated impacts, such as the affected populations, the dose received by the most exposed individual, the overall radiological impact and the doses to the public en route and the possible detriment to their health. In Spain a long “radioactive route” of more than 2,000 kilometres is expected, along which spent nuclear fuels will be transported – foreseeably by road – from the nuclear power plants to the CTS facility. The route will also be used for the vitrified wastes from fuel reprocessing of the fuel from Vandellós I nuclear power plant, which are currently in France. In conclusion, it may be observed that the emission of ionising radiations from transport of high level radioactive wastes in Spain is insignificant, in normal operations, as regards the generation of adverse effects for human health, and that the radiological impact may be considered negligible. In the event of an accident, the possibility of which is remote, the emissions will not be also a very determining factor as regards adverse effects for human health.

Efficient algorithms for dynamic probabilistic safety assessment. An application to convex damage domain.

The design of nuclear power plant has to follow a number of regulations aimed at limiting the risks inherent in this type of installation. The goal is to prevent and to limit the consequences of any possible incident that might threaten the public or the environment. To verify that the safety requirements are met a safety assessment process is followed. Safety analysis is as key component of a safety assessment, which incorporates both probabilistic and deterministic approaches. The deterministic approach attempts to ensure that the various situations, and in particular accidents, that are considered to be plausible, have been taken into account, and that the monitoring systems and engineered safety and safeguard systems will be capable of ensuring the safety goals. On the other hand, probabilistic safety analysis tries to demonstrate that the safety requirements are met for potential accidents both within and beyond the design basis, thus identifying vulnerabilities not necessarily accessible through deterministic safety analysis alone. Probabilistic safety assessment (PSA) methodology is widely used in the nuclear industry and is especially effective in comprehensive assessment of the measures needed to prevent accidents with small probability but severe consequences. Still, the trend towards a risk informed regulation (RIR) demanded a more extended use of risk assessment techniques with a significant need to further extend PSA’s scope and quality. Here is where the theory of stimulated dynamics (TSD) intervenes, as it is the mathematical foundation of the integrated safety assessment (ISA) methodology developed by the CSN(Consejo de Seguridad Nuclear) branch of Modelling and Simulation (MOSI). Such methodology attempts to extend classical PSA including accident dynamic analysis, an assessment of the damage associated to the transients and a computation of the damage frequency. The application of this ISA methodology requires a computational framework called SCAIS (Simulation Code System for Integrated Safety Assessment). SCAIS provides accident dynamic analysis support through simulation of nuclear accident sequences and operating procedures. Furthermore, it includes probabilistic quantification of fault trees and sequences; and integration and statistic treatment of risk metrics. SCAIS comprehensively implies an intensive use of code coupling techniques to join typical thermal hydraulic analysis, severe accident and probability calculation codes. The integration of accident simulation in the risk assessment process and thus requiring the use of complex nuclear plant models is what makes it so powerful, yet at the cost of an enormous increase in complexity. As the complexity of the process is primarily focused on such accident simulation codes, the question of whether it is possible to reduce the number of required simulation arises, which will be the focus of the present work. This document presents the work done on the investigation of more efficient techniques applied to the process of risk assessment inside the mentioned ISA methodology. Therefore such techniques will have the primary goal of decreasing the number of simulation needed for an adequate estimation of the damage probability. As the methodology and tools are relatively recent, there is not much work done inside this line of investigation, making it a quite difficult but necessary task, and because of time limitations the scope of the work had to be reduced. Therefore, some assumptions were made to work in simplified scenarios best suited for an initial approximation to the problem. The following section tries to explain in detail the process followed to design and test the developed techniques. Then, the next section introduces the general concepts and formulae of the TSD theory which are at the core of the risk assessment process. Afterwards a description of the simulation framework requirements and design is given. Followed by an introduction to the developed techniques, giving full detail of its mathematical background and its procedures. Later, the test case used is described and result from the application of the techniques is shown. Finally the conclusions are presented and future lines of work are exposed.

Estudio del módulo icrp del sistema jrodos de ayuda a la toma de decisiones en emergencias nucleares y radiológicas

Este proyecto se basa en el sistema JRodos de ayuda a la toma de decisiones en tiempo real en caso de emergencias nucleares y radiológicas. Tras una breve descripción del mismo, se presentan los modelos de cálculo que utiliza el sistema y la organización modular en la que se estructura el programa. Concretamente este documento se centra en un módulo desarrollado recientemente denominado ICRP y caracterizado por tener en cuenta todas las vías de exposición a la contaminación radiológica, incluida la vía de la ingestión que no se había tenido en cuenta en los módulos previos. Este modelo nuevo utiliza resultados obtenidos a partir de la cadena de escala local LSMC como datos de entrada, por lo que se lleva a cabo una descripción detalla del funcionamiento y de la ejecución tanto del módulo ICRP como de la cadena previa LSMC. Finalmente, se ejecuta un ejercicio ICRP usando los datos meteorológicos y de término fuentes reales que se utilizaron en el simulacro CURIEX 2013 realizado en el mes de noviembre de 2013 en la Central Nuclear de Almaraz. Se presenta paso a paso la ejecución de este ejercicio y posteriormente se analizan y explican los resultados obtenidos acompañados de elementos visuales proporcionados por el programa. This project is based on the real time online decision support system for nuclear emergency management called JRodos. After a brief description of it, the calculation models used by the system and its modular organization are presented. In particular, this paper focuses on a newly developed module named ICRP. This module is characterized by the consideration of the fact that all terrestrial exposure pathways, including ingestion, which has not been considered in previous modules. This new model uses the results obtained in a previous local scale model chain called LSMC as input. In this document a detailed description of the operation and implementation of both the ICRP module and its previous LSMC chain is presented. To conclude, an ICRP exercise is performed with real meteorological and source term data used in the simulation exercise CURIEX 2013 carried out in the Almaraz Nuclear Power Plant in November 2013. A stepwise realization of this exercise is presented and subsequently the results are deeply explained and analyzed supplemented with illustrations provided by the program.