PATRICIA DA SILVA PAGETTI DE OLIVEIRA

PATRICIA DA SILVA PAGETTI DE OLIVEIRA

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Development of the reliability assurance program in a Brazilian nuclear power plant subsidized by a reliability, availability and maintainability model
2023 - GOMES, J.M.; NETO, M.M.; MATURANA, M.C.; OLIVEIRA, P.S.P.
The main objective of this work is to present a methodology for the development of a Reliability Assurance Program (RAP) specific to a PWR experimental nuclear installation, through the analysis of the installation and the development of a preliminary RAP subsidized by a Reliability, Availability and Maintainability (RAM) model. The study of an evaluation was carried out in the long-term decay heat removal of the studied experimental plant, whose data were used for application of the RAP. The necessary steps for applying the developed RAP are followed, using the data from the assessment of the studied plant, resulting in a list of components of significant risk for the Program, and in the following steps of sending the list to the experts panel, ranking of SSCs by the panel and development of the final list of significant risk SSCs for using the list in the optimization of the plant. The RAP subsidized by a RAM model will be able to work with the logical relationships between each component of the plant for their effects on energy generation and with the quantitative prediction of the magnitude of each contributor to the occurrence of high-level events, and the developed methodology can be applicable throughout the experimental plant. In this way, it will be possible to implement the RAP in the plant, which will provide a structured way to meet the regulatory requirements for its licensing. Also, it will be possible to complement the plant safety analysis report, which must contain the RAP.
Risk-based design of electric power systems for non-conventional nuclear facilities at shutdown modes
2022 - BORSOI, S.S.; BARONI, D.B.; MATTAR NETO, M.; OLIVEIRA, P.S.P.; MATURANA, M.C.
The work presents a methodology for assessing the safety of electrical system designs for non-conventional nuclear facilities in shutdown. The methodology adopts the core damage frequency as the main risk measure to assess the different architectures of power systems in a non-conventional nuclear facility. Among the reasons is the absence of a specific regulatory basis for this type of installation. The adoption of standards for nuclear power plants by non-conventional nuclear facilities does not take into account the functional and operational particularities of these installations, imposing criteria that are often overestimated, which can even lead to an increase in the financial risk for carrying out the projects. Safety probabilistic analyzes become essential tools for the facilities design and licensing. The modeling and quantification of systems failures in charge of ensuring the nuclear safety of non-conventional nuclear facilities are carried out in the CAFTA software environment. In these studies, the analysis of electrical system configurations and their influence on the overall risk of the installation stand out.
External Events PSA
2022 - SILVA, T.P.; MATURANA, M.C.; OLIVEIRA, P.S.P. de; MATTAR NETO, M.
Since the Fukushima Daiichi accident, external events analysis has become a priority issue within regulatory bodies, operators, and designers, raising concerns about the capabilities of nuclear power plants to withstand severe conditions. Generally, the methodology applied to the Probabilistic Safety Assessment (PSA) of external events consists of the identification of potential single and combined external hazards, screening of external hazards, analysis of site and plant response, analysis of initiating events and quantification of accident sequences probabilities. Therefore, in this paper, the requirements and other information on new nuclear installations projects necessary to implement a comprehensive PSA of external events throughout plant lifetime are evaluated. In addition, it is necessary to clearly identify all the resources that must be available to continuously expand PSA scope to include all types of initiating events, levels of analysis and plant operation modes.
Licensing approach applicable to land facilities supporting nuclear-powered submarines
2022 - BARONI, D.B.; BORSOI, S.S.; MATTAR NETO, M.; OLIVEIRA, P.S.P.; MATURANA, M.C.
The nuclear licensing process is a fundamental stage for the design and deployment of a nuclear facility. In Brazil, the licensing process of Central Nuclear Almirante Álvaro Alberto (CNAAA) nuclear power plants, in Angra dos Reis - RJ, was established mainly based on the U.S. Nuclear Regulatory Commission (U.S.NRC) guidelines. However, for each purpose specific requirements are established which promote a standardization appropriate to the type of installation in question. Thus, not every nuclear installation can be adequately framed in the standards and requirements established for the licensing of a nuclear power plant, especially when considering nuclear facilities for strategic and defense purposes. For instance, the Specialized Maintenance Complex (CME) project is being developed by the Brazilian Navy and aims to offer all the structures and systems for support on land to the first Brazilian nuclear-powered submarine. Therefore, when considering the interfaces between maritime/naval systems and operations, the purpose and specificity of installations such as CME extrapolate the commonly established nuclear normative framework. Due to the innovation of this type of installation in Brazil, there is no specific regulation for its licensing, constituting a unique situation for both the Brazilian Navy (applicant) and the National Nuclear Energy Commission - CNEN (Brazilian Nuclear Licensing Agency, which, soon, will have its function incorporated into the National Nuclear Safety Authority, ANSN). Even when researching standards and other guides in ostensible sources of nations that hold nuclear reactor technology for naval propulsion (and land support facilities), no normative guidance dealing specifically with the safety analysis and licensing of this type of installation has been identified. Thus, this paper proposes a first approach and analysis of the standards used by the U.S. Department of Defense (U.S.DOE) comparing them to the standards of the U.S. Nuclear Regulatory Commission (U.S.NRC) aiming to compose a specific normative proposition to carry out the safety analysis and licensing of a nuclear-powered submarines land support facility.
Evaluation of “Safety Related” and “Important to Safety” terminology for safety classification of nuclear installation items in Brazil
2022 - BARONI, D.B.; BORSOI, S.S.; MATTAR NETO, M.; OLIVEIRA, P.S.P.; MATURANA, M.C.
In general terms, safety demonstration of nuclear installations is carried out through an assessment of compliance with design criteria and safety requirements established in national and international codes and standards applicable to each type of installation. In addition, a safety analysis consisting of installation behavior study during its useful lifetime, shall be developed considering normal operating conditions, transients, and postulated accidents, to determine safety margins and verify the adequacy of items designed to prevent accidents or mitigate their consequences. Also, design requirements applicable to each installation item depend on its classification with respect to safety. Thus, safety classification of structures, systems, and components (SSCs) must be performed based on adequate methods and clear and consistent criteria to ensure that an overall safety level expected for the installation is achieved. It is worth emphasizing the importance of the terminology adopted and the understanding of concepts definitions used in a safety classification process. The objective of this paper is to present a review of the application of “safety related item” and “item important to safety” terminology, evaluating definitions and interpretations given by the International Atomic Energy Agency (IAEA), the United States Nuclear Regulatory Commission (U.S.NRC) and the National Nuclear Energy Commission (CNEN) of Brazil. In this work, this subject is raised to demonstrate that divergent definitions and misinterpretations of concepts may result in inconsistencies in SSCs safety classification.
Risk-based design of electric power systems for non-conventional nuclear facilities at shutdown modes
2021 - BORSOI, S.S.; BARONI, D.B.; MATTAR NETO, M.; OLIVEIRA, P.S.P.; MATURANA, M.C.
External Events PSA
2021 - SILVA, T.P. da; MATURANA, M.C.; MATTAR NETO, M.; OLIVEIRA, P.S.P. de
Licensing approach for nuclear-powered submarines land support facilities
2021 - BARONI, D.B.; BORSOI, S.S.; MATTAR NETO, MIGUEL; OLIVEIRA, P.S.P. de; MATURANA, M.C.
Evaluation of “Safety Related” and “Important to Safety” terminology for safety classification of nuclear installation items in Brazil
2021 - BARONI, D.B.; BORSOI, S.S.; MATTAR NETO, M.; OLIVEIRA, P.S.P.; MATURANA, M.C.
Overview of seismic probabilistic safety assessment applied to a nuclear installation located in a low seismicity zone
2021 - OLIVEIRA, ELLISON A. de; OLIVEIRA, PATRICIA da S.P. de; MATTAR NETO, MIGUEL; MATURANA, MARCOS C.
Deterministic and probabilistic safety analysis methodologies have been developed and updated based on operational experience, investigation of past incidents or accidents, and analysis of postulated initiating events in nuclear plants in order to maintain the protection of workers, the public and the environment. The evaluation of accident sequences and the total radiological risk resulting from off-site releases are general objectives addressed by these methodologies. There are hazards that continually challenge the safety of a nuclear facility or its nearby area. In particular, seismic events represent a major contributor to the risk of a nuclear accident. Different levels of ground motion induced by earthquakes may be experienced by structures, systems and components (SSCs) of an installation. In this context, a seismic hazard analysis, seismic demand analysis and seismic fragility analysis must be carried out in order to characterize the local seismic hazard and seismic demands on SSCs, allowing an adequate seismic classification of SSCs, even for installations located in sites with low seismicity. In this article, a general description of the Seismic Probabilistic Safety Assessment (Seismic PSA) methodology is presented, emphasizing the supporting studies. This methodology shall be applied to an experimental nuclear installation containing a PWR reactor designed for naval propulsion to be installed in a low seismicity zone in Brazil.