O objetivo deste trabalho é apresentar a base de dados que foi desenvolvida para armazenar dados técnicos e processar dados sobre operação, falha e manutenção de equipamentos dos reatores nucleares de pesquisa localizados no Instituto de Pesquisas Energéticas e Nucleares (IPEN), em São Paulo - SP. Os dados extraídos desta base poderão ser aplicados na Análise Probabilística de Segurança dos reatores de pesquisa ou em avaliações quantitativas menos complexas relacionadas à segurança, confiabilidade, disponibilidade e manutenibilidade destas instalações. Esta base de dados foi desenvolvida de modo a permitir que as informações nela contidas estejam disponíveis aos usuários da rede corporativa, que é a intranet do IPEN. Os profissionais interessados deverão ser devidamente cadastrados pelo administrador do sistema, para que possam efetuar a consulta e/ou o manuseio dos dados. O modelo lógico e físico da base de dados foi representado por um diagrama de entidades e relacionamento e está de acordo com os módulos de segurança instalados na intranet do IPEN. O sistema de gerenciamento da base de dados foi desenvolvido com o MySQL, o qual utiliza a linguagem SQL como interface. A linguagem de programação PHP foi usada para permitir o manuseio da base de dados pelo usuário. Ao final deste trabalho, foi gerado um sistema de gerenciamento de base de dados capaz de fornecer as informações de modo otimizado e com bom desempenho. / The objective of this work is to present the computational database that was developed to store technical information and process data on component operation, failure and maintenance for the nuclear research reactors located at the Nuclear and Energy Research Institute (Instituto de Pesquisas Energéticas e Nucleares, IPEN), in São Paulo, Brazil. Data extracted from this database may be applied in the Probabilistic Safety Analysis of these research reactors or in less complex quantitative assessments related to safety, reliability, availability and maintainability of these facilities. This database may be accessed by users of the corporate network, named IPEN intranet. Professionals who require the access to the database must be duly registered by the system administrator, so that they will be able to consult and handle the information. The logical model adopted to represent the database structure is an entity-relationship model, which is in accordance with the protocols installed in IPEN intranet. The open-source relational database management system called MySQL, which is based on the Structured Query Language (SQL), was used in the development of this work. The PHP programming language was adopted to allow users to handle the database. Finally, the main result of this work was the creation a web application for the component reliability database named PSADB, specifically developed for the research reactors of IPEN; furthermore, the database management system provides relevant information efficiently.

O objetivo deste trabalho é apresentar a base de dados que foi desenvolvida para armazenar dados técnicos e processar dados sobre operação, falha e manutenção de equipamentos dos reatores nucleares de pesquisa localizados no Instituto de Pesquisas Energéticas e Nucleares (IPEN), em São Paulo - SP. Os dados extraídos desta base poderão ser aplicados na Análise Probabilística de Segurança dos reatores de pesquisa ou em avaliações quantitativas menos complexas relacionadas à segurança, confiabilidade, disponibilidade e manutenibilidade destas instalações. Esta base de dados foi desenvolvida de modo a permitir que as informações nela contidas estejam disponíveis aos usuários da rede corporativa, que é a intranet do IPEN. Os profissionais interessados deverão ser devidamente cadastrados pelo administrador do sistema, para que possam efetuar a consulta e/ou o manuseio dos dados. O modelo lógico e físico da base de dados foi representado por um diagrama de entidades e relacionamento e está de acordo com os módulos de segurança instalados na intranet do IPEN. O sistema de gerenciamento da base de dados foi desenvolvido com o MySQL, o qual utiliza a linguagem SQL como interface. A linguagem de programação PHP foi usada para permitir o manuseio da base de dados pelo usuário. Ao final deste trabalho, foi gerado um sistema de gerenciamento de base de dados capaz de fornecer as informações de modo otimizado e com bom desempenho. / The objective of this work is to present the computational database that was developed to store technical information and process data on component operation, failure and maintenance for the nuclear research reactors located at the Nuclear and Energy Research Institute (Instituto de Pesquisas Energéticas e Nucleares, IPEN), in São Paulo, Brazil. Data extracted from this database may be applied in the Probabilistic Safety Analysis of these research reactors or in less complex quantitative assessments related to safety, reliability, availability and maintainability of these facilities. This database may be accessed by users of the corporate network, named IPEN intranet. Professionals who require the access to the database must be duly registered by the system administrator, so that they will be able to consult and handle the information. The logical model adopted to represent the database structure is an entity-relationship model, which is in accordance with the protocols installed in IPEN intranet. The open-source relational database management system called MySQL, which is based on the Structured Query Language (SQL), was used in the development of this work. The PHP programming language was adopted to allow users to handle the database. Finally, the main result of this work was the creation a web application for the component reliability database named PSADB, specifically developed for the research reactors of IPEN; furthermore, the database management system provides relevant information efficiently.

There is so far no consensus on how to develop a reliability model of safety-related digitalinstrumentation and control (I&C) in a probabilistic safety assessment (PSA) of a nuclear powerplant. The objective of this thesis is to evaluate different approaches to model digital protectionsystems in a PSA for a nuclear power plant. This is accomplished by the development of a fault treemodel of the digital protection system for a fictive and simplified nuclear power plant, that act as areference model to be used for evaluation of different design alternatives and modelling principles.Common cause failures and spurious actuation signals are the major contributors to scenariosresulting in a core damage. A PSA model has to be sufficiently detailed in order for this to berepresented in the results. The impact on results such as core damage frequency and importance ofminimal cut sets from different fail-safe, voting logic and signal validation designs are significant,too. To further examine the differences between I&C designs and significance of different PSAmodelling solutions, the degree of realism of the example should be increased. This rapidlyintroduces complexity to the models resulting in a model that is more difficult to review and resultsthat is more difficult to interpret and even much simplified models tend to get rather complex.

<p>In the next few decades, the nuclear industry worldwide is expected to launch a set of reactors with advanced reactor designs. Generation-IV (GEN-IV) reactors are to display superior safety by incorporating additional passive safety concepts as well as improving accident management and minimization of consequences. Canada is in the early stages of conceiving its GEN-IV reactor design – the Supercritical Water Reactor (SCWR). The proposed design is based on the existing CANDU configurations and is expected to offer significant advances in thermal efficiency, fuel cycle sustainability, and relative cost of energy. Of particular interest is the reactor's ability to use inherent or passive safety concepts which will translate to the reactor being walk-away safe in an accident.</p> <p>Steam generators in CANDU remove decay heat by thermosyphoning in a loss of Class-IV power accident. This natural circulation process was a passive feature in GEN-II and GEN-III CANDUs. The SCWR's direct thermodynamic cycle implies steam generators are no longer incorporated into the design. This thesis examines how the SCWR compensates for the removal of a passive safety system element and the difference to the overall safety of the reactor following accidents. These results will be compared to the traditional CANDU's response in accidents to demonstrate the added value of this new reactor in maintaining the goal of no widespread core damage. Comparisons were also made between the SCWR and similar GEN-IV reactors in terms of design and response to various initiating events.</p> <p>Probabilistic Risk Analysis is used in this thesis to assess the SCWR design options. Although the SCWR is in the pre-conceptual design phase, the results of such risk assessment studies could affect the design, operation, and licensing of this new reactor. Future studies can build on this work to conduct more detailed analyses to characterise the SCWR's safety and reliability.</p> / Master of Applied Science (MASc)