PATRICIA DA SILVA PAGETTI DE OLIVEIRA

PATRICIA DA SILVA PAGETTI DE OLIVEIRA

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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.
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.