Desarrollo y aplicación de herramientas de valoración de riesgos tecnológicos en centrales nucleares españolas a partir de la técnica de "Análisis Probabilista de Seguridad"

The PhD thesis "Desarrollo y aplicación de herramientas de valoración de riesgos tecnológicos en centrales nucleares españolas a partir de la técnica de Análisis Probabilista de Seguridad” results from the research projects carried out between the Nuclear Engineering Research group, which belongs to UPC’s Divisió d’Enginyeria Nuclear, and one of the operator companies of Spanish nuclear power plants during the period of 2013-2017. Both the Secretaria de Recerca i Universitats of the Generalitat de Catalunya and the Ministerio de Econom ía, Industria y Competitividad of the Spanish government are also participants because they have awarded grants for the development of the projects herein presented. The main objective of this PhD thesis is to delve into the field of the risk-informed applications used in nuclear power plants, putting forward, developing, and testing new tools that allow a more detailed awareness and evaluation of nuclear power plants risks so that nuclear power plants can be operation and maintenance safely. For that purpose, the first part of this thesis presents the state of the art of risk-informed applications and their role in the frame of nuclear safety. The second part of the thesis presents the development and results analysis of a pilot probabilistic safety assessment for an independent spent fuel storage installation, which is devoted to store nuclear spent fuel. This assessment, which is pioneer in the Spanish nuclear industry, analyzes the installation’s response to abnormal events, and allows to quantify the risk related to the occurrence of an undesired consequence in this installation. Moreover, this part of the thesis presents a novel contribution to the field of human reliability analysis as applied to probabilistic safety assessment in the form of the development of a human reliability analysis methodology applicable to the storage installation handling operations. It is concluded that the contribution of the storage installation to the risk of radionuclide emission of the whole plant is virtually nonexistent. The third part of the thesis presents the development and application of methodologies and tools derived from fire probabilistic safety assessments whose purpose is to introduce the evaluation of the fire induced risk in the decision-making processes devoted to the configuration of systems, structures, and components in nuclear power plants. Historically, the assessment of the fire induced risk had been either carried out qualitatively or belittled mainly because of the late evolution of the fire induced risk quantitative assessment techniques. In consequence, the tools presented in this thesis are a novel contribution to the field of risk-informed decision making since they provide a quantitative evaluation of the fire induced risk by means of a fire probabilistic safety assessment. Two matrix-shaped novel tools are described and applied to the real case of a Spanish nuclear power plant. The use of these tools demonstrates that number of elements who are significant to the fire induced risk of the plant is small. The projects presented in this PhD thesis as well as others developed through the period 2013-2017 have presented in national and international conferences and have been published in international indexed journals.La tesi doctoral "Desarrollo y aplicación de herramientas de valoración de riesgos tecnológicos en centrales nucleares españolas a partir de la técnica de Análisis Probabilista de Seguridad" es producte dels projectes d'investigació portats a terme pel Nuclear Engineering Research Group, que pertany a la Divisió d'Enginyeria Nuclear de l'UPC, i una de les empreses operadores de centrals nuclears espanyoles durant el període 2013-2017. També són partícips d'aquesta tesi doctoral tant la Secretaria de Recerca i Universitats de la Generalitat de Catalunya com el Ministerio de Economía, Industria y Competitividad del Gobierno de España per prestar financiació per al desenvolupament dels treballs aquí exposats. El principal objectiu d'aquesta tesi doctoral és el de profunditzar al camp de les aplicacions informades pel risc utilitzades a centrals nuclears, proposant, desenvolupant, i provant noves eines que permetin conèixer i valorar amb més detall els riscos associats a les mateixes per a que les centrals nuclears puguin seguir sent operades i mantingudes de forma segura. Per donar compliment a aquest objectiu, la primera part de la tesi presenta l’estat de l’art de les aplicacions informades pel risc i la seva implicació en el marc de la seguretat nuclear. La segona part de la tesi presenta el desenvolupament i anàlisi de resultats d’un model pilot d’anàlisi probabilista de seguretat d’un magatzem temporal individualitzat, que es tracta d’una instal·lació d’emmagatzematge de combustible nuclear gastat. Aquest model, que es pioner en el marc de la industria nuclear espanyola, permet analitzar la resposta de la instal·lació davant de successos anormals, i permet quantificar el risc associat a la ocurrència d’una conseqüència no desitjada en aquesta instal·lació. A més, en una contribució novell al camp de l’anàlisi de fiabilitat humana per l’anàlisi probabilista de seguretat, aquesta part de la tesi doctoral presenta el desenvolupament d’una metodologia d’anàlisi de fiabilitat humana que és aplicable a les operacions a portar a terme al context concret d’un magatzem temporal individualitzat. En aquesta part de la tesi es conclou que la contribució del risc del magatzem temporal individualitzat al total del risc de la central nuclear és pràcticament nul·la. La tercera part de la tesi presenta el desenvolupament i aplicació de metodologies i eines, generades mitjançant un anàlisi probabilista de seguretat d’incendis, per incorporar el risc induït per incendis a les pràctiques habituals de presa de decisions al respecte de configuracions de sistemes, components, i estructures, a centrals nuclears. Històricament, la valoració del risc induït per incendis s’ha portat a terme mitjançant metodologies qualitatives o s’ha menyspreat a causa de, principalment, l’evolució tardana de les metodologies quantitatives d’anàlisi de riscos induïts per incendis. Per tant, les eines i metodologies presentades són una contribució novell al camp de la presa de decisions informada pel risc doncs proporcionen una valoració quantitativa del risc d’incendis mitjançant l’eina que suposa l’APS d’incendis. Es proposen i descriuen dues eines novells, en forma de matriu, que s’apliquen al cas real d’una central nuclear espanyola. L’aplicació d’aquestes eines ha conclòs que tot element amb impacte sobre el risc induït per incendis d’aquesta central està localitzat en un grup reduït d’equips, sistemes contra incendis, i zones d’incendi. Els treballs presentats en aquesta tesi doctoral, així com altres desenvolupats durant el període 2013-2017, han estat difosos a diversos congressos nacionals e internacionals, i han estat motiu d’articles de revista internacional i indexada.Postprint (published version

Development and assessment of fire-related risk unavailability matrices to support the application of the maintenance rule in a PWR nuclear power plant

Two methods are presented which serve to incorporate the fire-related risk into the current practices in nuclear power plants with respect to the assessment of configurations. The development of these methods is restricted to the compulsory use of fire probabilistic safety assessment (PSA) models. The first method is a fire protection systems and key safety functions unavailability matrix which is developed to identify structures, systems, and components significant for fire-related risk. The second method is a fire zones and key safety functions (KSFs) fire risk matrix which is useful to identify fire zones which are candidates for risk management actions. Specific selection and quantification methodologies have been developed to obtain the matrices. The Monte Carlo method has been used to assess the uncertainty of the unavailability matrix. The analysis shows that the uncertainty is sufficiently bounded. The significant fire-related risk is localized in six KSF representative components and one fire protection system which should be included in the maintenance rule. The unavailability of fire protection systems does not significantly affect the risk. The fire risk matrix identifies the fire zones that contribute the most to the fire-related risk. These zones belong to the control building and electric penetrations building.Peer ReviewedPostprint (published version

Equipment for the continuous measurement and identification of gamma radioactivity on aerosols

© 20xx IEEE. Personal use of this material is permitted. Permission from IEEE must be obtained for all other uses, in any current or future media, including reprinting/republishing this material for advertising or promotional purposes, creating new collective works, for resale or redistribution to servers or lists, or reuse of any copyrighted component of this work in other works.This paper describes an equipment for continuous measurement and identification of gamma radioactivity in aerosols developed by the Nuclear Engineering Research Group (NERG) at the Technical University of Catalonia (UPC) and Raditel Serveis i Subministraments Tecnològics, Ltd. A spectrometric analysis code has been specially designed for it. Spectrum analysis identifies and determines activity concentration of aerosol emitters captured by a fiberglass paper filter. This new equipment is currently operating in three radioactivity monitoring stations of the Environmental Radiological Surveillance Network of the Generalitat of Catalunya (local Catalan government): two near Ascó and Vandellòs Nuclear Power Plants in the province of Tarragona and one in the city of Barcelona. Two more monitors are expected to be deployed at Roses, Girona, Spain, and Puigcerdà, Barcelona, Spain. Measurements and evolution analysis results of emitters identified at these stations were also provided.Postprint (author's final draft

A new code for spectrometric analysis for environmental radiological surveillance on monitors focused on gamma radioactivity on aerosols

© 20xx IEEE. Personal use of this material is permitted. Permission from IEEE must be obtained for all other uses, in any current or future media, including reprinting/republishing this material for advertising or promotional purposes, creating new collective works, for resale or redistribution to servers or lists, or reuse of any copyrighted component of this work in other works.This paper presents a new code for the analysis of gamma spectra generated by an equipment for continuous measurement of gamma radioactivity in aerosols with paper ¿lter. It is called pGamma and has been developed by the Nuclear Engineering Research Group at the Technical University of Catalonia - Barcelona Tech and by Raditel Serveis i Subministraments Tecnològics, Ltd. The code has been developed to identify the gamma emitters and to determine their activity concentration. It generates alarms depending on the activity of the emitters and elaborates reports. Therefore it includes a library with NORM and arti¿cial emitters of interest. The code is being adapted to the monitors of the Environmental Radiological Surveillance Network of the local Catalan Government in Spain (Generalitat de Catalunya) and is used at three stations of the Network.Postprint (author's final draft

Fertilizer-drawn forward osmosis as a solution to improve the quality of wastewater treatment plant effluents used for agricultural irrigation

This study assesses the effectiveness of a forward osmosis (FO) pilot plant (max. Flow rate 0.36 m3/h) to improve the quality of a wastewater treatment plant (WWTP) effluent and its use for irrigating lettuce in greenhouse conditions. FO treated WWTP-effluent had nutrient levels comparable to well-water (control), except for ammonia and phosphates -which were recovered from the effluent- and potassium, which was added as fertilizer. FO also removed >95 % of organic pollutants from WWTP effluent. In addition, the toxicity associated with the WWTP effluent was reduced after FO treatment, reaching eleutheroembryo toxicity-biomarker levels similar to those found in well water. Finally, FO reduced the bacterial load of WWTP effluent, and significantly reduced its levels of ARGs. Lettuces irrigated with the three water sources did not show differences in terms of agronomic parameters, ARG and pathogen levels, and only some patterns were observed on plant metabolomics or transcriptomics. However, the potential accumulation of pollutants in the soil over time could amplify the impact of using WWTP effluent in crops compared to the use of FO-treated WWTP effluent.This study conducted in the frame of the project “Decision support-based approach for sustainable water reuse application in agricultural production (DSWAP)” which was funded from the Partnership on Research and Innovation in the Mediterranean Area (PRIMA) under grant agreement No 1822 . PRIMA is supported by the European Union‘s Horizon 2020. Mònica Escolà Casas acknowledges the Beatriu de Pinós 2018 grant-programme (MSCA grant agreement number 801370 ) for the funding. Authors are also grateful for the support of the Spanish Ministry of Science, Innovation, and University (MCIN/AEI/ 10.13039/501100011033 , grant RTI2018-096175-B-I00 ), and the Generalitat de Catalunya ( 2017SGR902 ). IDAEA-CSIC is a Centre of Excellence Severo Ochoa (Spanish Ministry of Science and Innovation, Project CEX2018-000794-S, ERDF A way of making Europe).Peer reviewe

Colombian consensus recommendations for diagnosis, management and treatment of the infection by SARS-COV-2/ COVID-19 in health care facilities - Recommendations from expert´s group based and informed on evidence

La Asociación Colombiana de Infectología (ACIN) y el Instituto de Evaluación de Nuevas Tecnologías de la Salud (IETS) conformó un grupo de trabajo para desarrollar recomendaciones informadas y basadas en evidencia, por consenso de expertos para la atención, diagnóstico y manejo de casos de Covid 19. Estas guías son dirigidas al personal de salud y buscar dar recomendaciones en los ámbitos de la atención en salud de los casos de Covid-19, en el contexto nacional de Colombia

Assessment and application of Human Reliability Aanalysis to an Independent Spent Fuel Storage Installation Probabilistic Safety Assessment

This Master Thesis is framed within a collaboration agreement between the Nuclear Engineering Research Group (NERG) of Universitat Politècnica de Catalunya (UPC) and a Spanish Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP). The main objective of the collaboration is to study and apply the Probabilistic Safety Assessment (PSA) methodology to risk-informed decision making. This study is part of the NPP Independent Spent Fuel Storage Installation (ISFSI) PSA requested by the Spanish NPP. It is the continuation of the final degree Project “Estudio piloto para el análisis del riesgo asociado a un Almacén Temporal Individualizado. Aplicación de la metodología APS” [1], which developed a pilot ISFSI’s PSA model without Human Reliability Analysis (HRA). The objective of this thesis is to apply a HRA to the Spanish NPP Independent Spent Fuel Storage Installation and then implement the HRA results into the ISFSI’s PSA model in order to evaluate the impact of human performance on the ISFSI’s Risk. In consequence, the project is divided in two different parts, the HRA development and the HRA implementation into the PSA model. The first part is more research related. On the other hand, the second part is more engineering related. The ISFSI’s HRA application is based on the methodology described in the regulatory guide NUREG-1880, “ATHEANA User’s Guide Final Report”, as recommended by several Nuclear Regulatory Commission (NRC) publications [2][3]. This methodology requires the participation of ISFSI’s Subject-Matter Experts (SMEs). SMEs are not available since the Spanish NPP has little experience in ISFSI operations. Therefore, it has been decided to develop an HRA methodology which does not need SMEs to be carried out. To do so, the contribution of the SME’s has been replaced with the use of other HRA methodologies, namely THERP and SPAR-H. In consequence, an experimental hybrid ATHEANA-based HRA methodology has been used to perform the analysis. The HRA results should be considered illustrative rather than definitive since several assumptions have been taken to apply the methodology and describe human actions. Furthermore, the HRA results cannot be compared with Nuclear Industry data since no ATHEANA-based ISFSI HRA has been published yet. The most important operations from a Risk point of view [1] are the human actions performed inside the Spent Fuel Storage Building (SFSB). Therefore, the HRA has been limited to the analysis of these human actions. Some examples are loading the canister with wrong Spent Fuel Elements (SFEs), also known as Misload, and different canister drop scenarios. The introduction of the HRA results as Human Error Probabilities into the PSA model implies the modification of the Initiating Events (IEs) treatment. Fault Tree (FT) models have been Pág. 2 Report developed to include both Human Failure Events (HFEs) and Crane components failure into the assessment of IEs frequency. These models have replaced all the Initiating Events previously included in the PSA model. The comparison of the PSA model results, with and without HRA, in terms of Risk yields that human performance has a substantial impact on the ISFSI’s Risk. Namely, the implementation of the HFEs increases the overall ISFSI’s Risk roughly 2 orders of magnitude. However, the results obtained in this thesis are illustrative since there are many uncertainties and assumptions in the HRA analysis. It cannot be directly concluded that human performance will actually have an important impact in the Risk. However, it is recommended to take into consideration that human performance could be of importance and that further studies should be carried out

Desarrollo y aplicación de herramientas de valoración de riesgos tecnológicos en centrales nucleares españolas a partir de la técnica de "Análisis Probabilista de Seguridad"

The PhD thesis "Desarrollo y aplicación de herramientas de valoración de riesgos tecnológicos en centrales nucleares españolas a partir de la técnica de Análisis Probabilista de Seguridad” results from the research projects carried out between the Nuclear Engineering Research group, which belongs to UPC’s Divisió d’Enginyeria Nuclear, and one of the operator companies of Spanish nuclear power plants during the period of 2013-2017. Both the Secretaria de Recerca i Universitats of the Generalitat de Catalunya and the Ministerio de Econom ía, Industria y Competitividad of the Spanish government are also participants because they have awarded grants for the development of the projects herein presented. The main objective of this PhD thesis is to delve into the field of the risk-informed applications used in nuclear power plants, putting forward, developing, and testing new tools that allow a more detailed awareness and evaluation of nuclear power plants risks so that nuclear power plants can be operation and maintenance safely. For that purpose, the first part of this thesis presents the state of the art of risk-informed applications and their role in the frame of nuclear safety. The second part of the thesis presents the development and results analysis of a pilot probabilistic safety assessment for an independent spent fuel storage installation, which is devoted to store nuclear spent fuel. This assessment, which is pioneer in the Spanish nuclear industry, analyzes the installation’s response to abnormal events, and allows to quantify the risk related to the occurrence of an undesired consequence in this installation. Moreover, this part of the thesis presents a novel contribution to the field of human reliability analysis as applied to probabilistic safety assessment in the form of the development of a human reliability analysis methodology applicable to the storage installation handling operations. It is concluded that the contribution of the storage installation to the risk of radionuclide emission of the whole plant is virtually nonexistent. The third part of the thesis presents the development and application of methodologies and tools derived from fire probabilistic safety assessments whose purpose is to introduce the evaluation of the fire induced risk in the decision-making processes devoted to the configuration of systems, structures, and components in nuclear power plants. Historically, the assessment of the fire induced risk had been either carried out qualitatively or belittled mainly because of the late evolution of the fire induced risk quantitative assessment techniques. In consequence, the tools presented in this thesis are a novel contribution to the field of risk-informed decision making since they provide a quantitative evaluation of the fire induced risk by means of a fire probabilistic safety assessment. Two matrix-shaped novel tools are described and applied to the real case of a Spanish nuclear power plant. The use of these tools demonstrates that number of elements who are significant to the fire induced risk of the plant is small. The projects presented in this PhD thesis as well as others developed through the period 2013-2017 have presented in national and international conferences and have been published in international indexed journals.La tesi doctoral "Desarrollo y aplicación de herramientas de valoración de riesgos tecnológicos en centrales nucleares españolas a partir de la técnica de Análisis Probabilista de Seguridad" es producte dels projectes d'investigació portats a terme pel Nuclear Engineering Research Group, que pertany a la Divisió d'Enginyeria Nuclear de l'UPC, i una de les empreses operadores de centrals nuclears espanyoles durant el període 2013-2017. També són partícips d'aquesta tesi doctoral tant la Secretaria de Recerca i Universitats de la Generalitat de Catalunya com el Ministerio de Economía, Industria y Competitividad del Gobierno de España per prestar financiació per al desenvolupament dels treballs aquí exposats. El principal objectiu d'aquesta tesi doctoral és el de profunditzar al camp de les aplicacions informades pel risc utilitzades a centrals nuclears, proposant, desenvolupant, i provant noves eines que permetin conèixer i valorar amb més detall els riscos associats a les mateixes per a que les centrals nuclears puguin seguir sent operades i mantingudes de forma segura. Per donar compliment a aquest objectiu, la primera part de la tesi presenta l’estat de l’art de les aplicacions informades pel risc i la seva implicació en el marc de la seguretat nuclear. La segona part de la tesi presenta el desenvolupament i anàlisi de resultats d’un model pilot d’anàlisi probabilista de seguretat d’un magatzem temporal individualitzat, que es tracta d’una instal·lació d’emmagatzematge de combustible nuclear gastat. Aquest model, que es pioner en el marc de la industria nuclear espanyola, permet analitzar la resposta de la instal·lació davant de successos anormals, i permet quantificar el risc associat a la ocurrència d’una conseqüència no desitjada en aquesta instal·lació. A més, en una contribució novell al camp de l’anàlisi de fiabilitat humana per l’anàlisi probabilista de seguretat, aquesta part de la tesi doctoral presenta el desenvolupament d’una metodologia d’anàlisi de fiabilitat humana que és aplicable a les operacions a portar a terme al context concret d’un magatzem temporal individualitzat. En aquesta part de la tesi es conclou que la contribució del risc del magatzem temporal individualitzat al total del risc de la central nuclear és pràcticament nul·la. La tercera part de la tesi presenta el desenvolupament i aplicació de metodologies i eines, generades mitjançant un anàlisi probabilista de seguretat d’incendis, per incorporar el risc induït per incendis a les pràctiques habituals de presa de decisions al respecte de configuracions de sistemes, components, i estructures, a centrals nuclears. Històricament, la valoració del risc induït per incendis s’ha portat a terme mitjançant metodologies qualitatives o s’ha menyspreat a causa de, principalment, l’evolució tardana de les metodologies quantitatives d’anàlisi de riscos induïts per incendis. Per tant, les eines i metodologies presentades són una contribució novell al camp de la presa de decisions informada pel risc doncs proporcionen una valoració quantitativa del risc d’incendis mitjançant l’eina que suposa l’APS d’incendis. Es proposen i descriuen dues eines novells, en forma de matriu, que s’apliquen al cas real d’una central nuclear espanyola. L’aplicació d’aquestes eines ha conclòs que tot element amb impacte sobre el risc induït per incendis d’aquesta central està localitzat en un grup reduït d’equips, sistemes contra incendis, i zones d’incendi. Els treballs presentats en aquesta tesi doctoral, així com altres desenvolupats durant el període 2013-2017, han estat difosos a diversos congressos nacionals e internacionals, i han estat motiu d’articles de revista internacional i indexada

Assessment and application of Human Reliability Aanalysis to an Independent Spent Fuel Storage Installation Probabilistic Safety Assessment

This Master Thesis is framed within a collaboration agreement between the Nuclear Engineering Research Group (NERG) of Universitat Politècnica de Catalunya (UPC) and a Spanish Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP). The main objective of the collaboration is to study and apply the Probabilistic Safety Assessment (PSA) methodology to risk-informed decision making. This study is part of the NPP Independent Spent Fuel Storage Installation (ISFSI) PSA requested by the Spanish NPP. It is the continuation of the final degree Project “Estudio piloto para el análisis del riesgo asociado a un Almacén Temporal Individualizado. Aplicación de la metodología APS” [1], which developed a pilot ISFSI’s PSA model without Human Reliability Analysis (HRA). The objective of this thesis is to apply a HRA to the Spanish NPP Independent Spent Fuel Storage Installation and then implement the HRA results into the ISFSI’s PSA model in order to evaluate the impact of human performance on the ISFSI’s Risk. In consequence, the project is divided in two different parts, the HRA development and the HRA implementation into the PSA model. The first part is more research related. On the other hand, the second part is more engineering related. The ISFSI’s HRA application is based on the methodology described in the regulatory guide NUREG-1880, “ATHEANA User’s Guide Final Report”, as recommended by several Nuclear Regulatory Commission (NRC) publications [2][3]. This methodology requires the participation of ISFSI’s Subject-Matter Experts (SMEs). SMEs are not available since the Spanish NPP has little experience in ISFSI operations. Therefore, it has been decided to develop an HRA methodology which does not need SMEs to be carried out. To do so, the contribution of the SME’s has been replaced with the use of other HRA methodologies, namely THERP and SPAR-H. In consequence, an experimental hybrid ATHEANA-based HRA methodology has been used to perform the analysis. The HRA results should be considered illustrative rather than definitive since several assumptions have been taken to apply the methodology and describe human actions. Furthermore, the HRA results cannot be compared with Nuclear Industry data since no ATHEANA-based ISFSI HRA has been published yet. The most important operations from a Risk point of view [1] are the human actions performed inside the Spent Fuel Storage Building (SFSB). Therefore, the HRA has been limited to the analysis of these human actions. Some examples are loading the canister with wrong Spent Fuel Elements (SFEs), also known as Misload, and different canister drop scenarios. The introduction of the HRA results as Human Error Probabilities into the PSA model implies the modification of the Initiating Events (IEs) treatment. Fault Tree (FT) models have been Pág. 2 Report developed to include both Human Failure Events (HFEs) and Crane components failure into the assessment of IEs frequency. These models have replaced all the Initiating Events previously included in the PSA model. The comparison of the PSA model results, with and without HRA, in terms of Risk yields that human performance has a substantial impact on the ISFSI’s Risk. Namely, the implementation of the HFEs increases the overall ISFSI’s Risk roughly 2 orders of magnitude. However, the results obtained in this thesis are illustrative since there are many uncertainties and assumptions in the HRA analysis. It cannot be directly concluded that human performance will actually have an important impact in the Risk. However, it is recommended to take into consideration that human performance could be of importance and that further studies should be carried out

Estudio piloto para el análisis del riesgo asociado a un Almacén Temporal Individualizado. Aplicación de la metodología APS

El Proyecto Final de Carrera “Estudio piloto para el análisis del riesgo asociado a un Almacén Temporal Individualizado. Aplicación de la metodología APS” se enmarca en un convenio de colaboración entre una central nuclear y el Departament de Física i Enginyeria Nuclear de la Universitat Politècnica de Catalunya (UPC) para la investigación y aplicación de los APS en decisiones basadas en el riesgo. El proyecto se basa en la metodología descrita en la guía reguladora NUREG/CR-2300, “PRA Procedures Guide” (editado por la Nuclear Regulatory Comission) y los documentos Probabilistic Risk Assessment (PRA) of Bolted Storage Casks: Updated Quantification and Analysis Report” (editado por EPRI) y “A Pilot Probabilistic Risk Assessment Of a Dry Cask Storage System At a Nuclear Power Plant” (editado por la Nuclear Regulatory Commission) y abarca las fases de un APS nivel 2. En primer lugar se ha realizado la familiarización con la instalación y con las técnicas probabilísticas a aplicar. Esta parte del proyecto incluye la descripción del Almacén Temporal Individualizado, la definición de las Funciones Clave de Seguridad (FCS), la identificación de los Sistemas encargados de cumplirlas, la definición de los criterios de éxito de éstos Sistemas y la identificación de los Sucesos Iniciadores y de los Escenarios. Posteriormente se han delineado las Secuencias de Accidente mediante la modelización de los Árboles de Eventos. Asimismo, se han establecido niveles de importancia discretos para el estado final de las Secuencias de Accidente, que consiste en la Liberación de Radionúclidos. Una vez delineadas las secuencias, se ha realizado un Análisis de Datos, el cual ha consistido en el cálculo de las frecuencias de los Sucesos Iniciadores y en la adecuación al procedimiento de manipulación de la central de resultados de análisis estructurales y termohidráulicos de contenedores utilizados en otras plantas. A continuación, se ha desarrollado el análisis del Sistema de Ventilación HVAC del Edificio de Combustible. Se ha modelizado el fallo del Sistema mediante un Árbol de Fallos en base a su funcionamiento en caso de accidente en el interior del Edificio de Combustible. El Sistema de Ventilación tiene una función de mitigación dentro de las Secuencias de Accidente. Finalmente, se ha realizado el proceso de Cuantificación mediante el software RiskSpectrum®. La Frecuencia de Liberación de Radionúclidos (FLR) del primer año obtenida es de 1,48E-08 (año·contenedor)-1, y en años venideros es de 6,68E-11 (año·contenedor)-1. Asimismo se ha analizado la magnitud de la Liberación de Radionúclidos para cada Secuencia de Accidente y se ha desarrollado el Análisis de Resultados, permitiendo extraer las pertinentes conclusiones y recomendaciones del estudio. Como conclusión final, se ha comprobado que el valor de FLR de primer año es coherente en comparación al de Almacenes Temporales Individualizados de otras centrales, este hecho le confiere validez al resultado