GAIANE SABUNDJIAN

Resumo

Possui graduação em Bacharel e Licenciatura Em Física pela Pontifícia Universidade Católica de São Paulo(1978), mestrado em Tecnologia Nuclear / Reatores de Potência pelo Instituto de Pesquisas Energéticas e Nucleares(1981) e doutorado em Engenharia Mecânica pela Escola Politêcnica de São Paulo(1999). Atualmente é TECNOLOGISTA SENIOR do Comissão Nacional de Energia Nuclear. Tem experiência na área de Engenharia Nuclear, com ênfase em Tecnologia dos Reatores. Atuando principalmente nos seguintes temas:Elementos Finitos, Formulação Petrov-Galerkin, Equações de Navier-Stokes, Fluidos Incompressíveis, Funções de Expansão Hierárquicas. (Texto extraído do Currículo Lattes em 13 out. 2021)

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  • Artigo IPEN-doc 30242
    RELAP5 code theoretical simulation of the experiment of natural circulation STAR
    2023 - SABUNDJIAN, GAIANE; BELCHIOR JUNIOR, ANTONIO; MAPRELIAN, EDUARDO
    Studies have been carried out on nuclear reactors with safety characteristics that do not depend on external intervention by operators or even on an external energy source. In this type of reactors, cooling is carried out by natural circulation, both during normal operation and during shutdown. For this reason, the STAR experiment was built in the IEA-R1 research reactor installed at the Institute for Energy and Nuclear Research (IPEN) – Brazil, with the aim of simulating experiments with the RELAP5 code in order to validate its models in two scenarios: total and partial emptying of the STAR. The results obtained with the RELAP5 code were compared to the experimental ones for the two proposed scenarios. These results showed that the mathematical correlations contained in RELAP5 are capable of reliably and safely reproducing the phenomenology of natural circulation in nuclear reactors.
  • Artigo IPEN-doc 25153
    Uncertainty calculation in small break LOCA in the emergency core cooling system connected to the hot leg of Angra 2 nuclear power plant
    2018 - BORGES, EDUARDO M.; SABUNDJIAN, GAIANE; DAURIA, FRANCESCO; PETRUZZI, ALESSANDRO
    Owing to the occurrence of nuclear accidents, worldwide nuclear regulatory organisations included the analysis of accidents considered as design basis accidents – Loss of Coolant Accident (large and small-break, LBLOCA or SBLOCA) – in the safety analysis reports of nuclear facilities. In Brazil, the tool selected by the licensing authority, Comissão Nacional de Energia Nuclear (CNEN), is RELAP5 Code. The aim of this paper is the evaluation of the performance of the Emergency Core Cooling System (ECCS) of Angra 2 nuclear reactor during SBLOCA. In this study, the RELAP5 code and the Code Internal Assessment of Uncertainty (CIAU) were used to simulate and analyse the uncertainties of the results. The postulated accident is the SBLOCA in the hot leg connected to the ECCS described in the Final Safety Analysis Report of Angra 2 (FSAR/A2). The results from this study were satisfactory when compared with the FSAR/A2.