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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Agora exibindo 1 - 10 de 35
  • Artigo IPEN-doc 29115
    Consequence analysis of a Station Blackout in Brazilian nuclear power plant Angra 2
    2022 - AGUIAR, A.S.; LEE, S.M.; SABUNDJIAN, G.
    The article consists, through a Severe Accident, evaluating the impact of radionuclides released into the atmosphere in the vicinity at Nuclear Power Plant. The source term used in present work is obtained by means of proportionality between Angra 1 and Angra 2. That is, the source term of Angra 2 is calculated based on its activity estimated from numbers of fuel pellets of both power plants and the already known activity of Angra 1. This calculation resulted in total activity of Angra 2 equivalent to 146.18% of activity of Angra 1. The results indicate that for severe accident scenarios, the protective measures to be adopted will be general emergency; and the impact area, which currently has a distance of 5 km, would become greater than this value.
  • Artigo IPEN-doc 27242
    Análise do acidente de perda de refrigerante primário devido a quebra da linha de surto do pressurizador da usina nuclear Angra 2
    2020 - BORGES, EDUARDO M.; CONTI, THADEU das N.; SANCHES, ANDREA; SABUNDJIAN, GAIANE
    O objetivo deste trabalho foi simular e avaliar com o código RELAP5 o acidente base de projeto de perda de refrigerante primário devido a uma ruptura média na linha de surto do pressurizador da usina nuclear Angra 2. Este acidente foi uma quebra do tipo guilhotina ou seja 100% na linha de surto do pressurizador, que representa uma ruptura de 437 cm². Nesta análise, verificou-se que as lógicas de atuação do Sistema de Proteção do Reator (SPR) e do Sistema de Resfriamento de Emergência do Núcleo (SREN) de Angra 2, utilizadas nesta simulação, funcionaram corretamente, mantendo a integridade do núcleo com as temperaturas do núcleo em níveis aceitáveis durante todo o evento. Os resultados obtidos foram satisfatórios, quando comparados com os apresentados no Relatório Final de Análise de Segurança de Angra 2 (FSAR/A2).
  • Artigo IPEN-doc 26388
    Analysis of the protective actions in the Emergency Planning Zones (EPZs) in the Angra dos Reis region through the calculation of the dose for public individuals due to a severe accident at the Angra 2 Nuclear Plant
    2019 - AGUIAR, ANDRE S.; LEE, SEUNG M.; SABUNDJIAN, G.
    This work presents the results of the computational simulations of the consequences of a severe accident in Angra 2 nuclear power plant. The severe accident was supposed to be caused by a rupture of 380cm2 in the primary reactor coolant system resulting in loss of coolant. Since the area of the rupture is quite smaller than the total flow area of the pipe of the primary coolant system, 4418cm2, the accident is classified as a small break loss of coolant accident. However, this rupture by itself would not bring the system about a severe accident, which must involve a considerable damage in the nuclear core. Thus, some boundary conditions were added to the problem in order to set a scenario of this kind of accident, which was simulated by means of the MELCOR code. The results obtained by this code show that the release of the radionuclide to the environment starts at the opening of the containment relief valve, and this valve, in turn, opens when the containment pressure reaches 7bar, at 168 hours after the break of the pipe of the coolant system, according to the simulation. The program used for calculation of the release of the radionuclides to the surrounding region of the nuclear plant was the CALMET/CALPUFF code, so that the atmospheric and transport model were elaborated for this code. A source term was used in order to carry out an analysis of the protective actions in the emergency planning zones by means dose calculation for individuals of the public, and it was based on two different scenarios: first scenario, release of the total activity to the atmosphere of Xe, Cs, Ba and Te, after 2h simulation and second scenario, release of the total activity to the atmosphere of Xe, Cs, Ba and Te, after 168h of simulation.
  • Artigo IPEN-doc 26382
    The cross sections obtained by the serpent code and formatting the input data for the PARCS code using the GenPMAXS code
    2019 - SANCHEZ, ANDREA; CARLUCCIO, THIAGO; SABUNDJIAN, GAIANE
    The Purdue Advanced Reactor Core Simulator (PARCS) is a computer code that solves the time-dependent two-group neutron diffusion equation in three-dimensional Cartesian geometry using nodal methods to obtain the transient neutron flux distribution. The code is used in the analysis of reactivity-initiated accidents in light-water reactors where spatial effects may be important. It may be run in the stand-alone mode or coupled to other NRC thermal-hydraulic codes such as RELAP5. The PARCS neutron code accepts libraries from HELIOS, TRITON, WIMS, SERPENT, etc., codes, but for some libraries is required special formatting. In the case of the SERPENT code, the GenPMAXS code must be used for the PARCS code to be able to read the cross sections library correctly. This work is part of a study on the PARCS/RELAP5 coupling for analyzing the control rod ejection of the Angra 2 reactor core. For this case, the core cross sections were obtained for 6 different branches varying the fuel temperature, moderator temperature, moderator density, boron concentration and considering rods removed and inserted. After obtaining the cross sections with the code SERPENT 2.1.26, these data were passed by a special formatting realized with the code GenPMAXS v6.2. Since GenPMAXS has several options controlling how to process the cross-sections generated by Serpent, a several doubts arose about the correct use of the code. When the doubts are answered, the file with the input data that will be used for the PARCS / RELAP coupling can be built.
  • Artigo IPEN-doc 26381
    Small break loss of coolant accident of 200 cm² in cold leg of primary loop of ANGRA 2 nuclear power reactor evaluation
    2019 - BORGES, EDUARDO M.; SABUNDJIAN, GAIANE
    The aim of this paper is evaluated the consequences to ANGRA 2 nuclear power reactor and to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 200cm2 of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. The results obtained for ANGRA 2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core.
  • Artigo IPEN-doc 26380
    RELAP5 code simulation of the small break loss of coolant accident of 80 cm² in the cold leg of Angra2 primary loop
    2019 - BORGES, EDUARDO M.; SABUNDJIAN, GAIANE; BRAZ FILHO, FRANCISCO A.; GUIMARÃES, LAMARTINE N.F.
    The aim of this paper was to simulate and evaluate the basic design accident of 80 cm² small break loss of coolant accident (SBLOCA) in the cold leg of the primary loop of the Angra2 nuclear power plant. In this simulation, it was verified that the actuation logics of the Angra2 Reactor Protection System (RPS) and the Emergency Core Cooling System (ECCS) used in this simulation worked correctly, maintaining core integrity with acceptable temperatures throughout the event. The results obtained were satisfactory when compared with those presented by the Angra2 Final Safety Analysis Report (FSAR/A2).
  • Artigo IPEN-doc 26345
    An Angra 2 LBLOCA simulation model for RELAP5MOD3.3 code with uncertainty analysis
    2019 - MADEIRA, ALZIRA A.; PEREIRA, LUIZ C.M.; SABUNDJIAN, GAIANE
    This paper describes the activities related to the work planned within Project BRA3.01/12 between CNEN and the European Community, relatively to its Task 2.1 (independent uncertainty quantification and sensitivity analysis utilizing the computational tool SUSA for the calculus related to LOCA simulation for licensing matter). SUSA software has been applied to the reference case, a double-ended LBLOCA in Angra 2, simulated with a RELAP5 code nodalization developed by the thermal hydraulic technicians of CNEN and its research institutes. This original nodalization has been improved for the development of the main objective of Task 2.1. The recommendations that our European counterparts provided on the last workshop, held at CNEN in Rio de Janeiro from January 28th to February 2nd, 2018, have been implemented as far as feasible.
  • Artigo IPEN-doc 25715
    Containment pressure analysis methodology during a LBLOCA with COCOSYS code
    2018 - SILVA, DAYANE F.; LIMA, ANA C. de S.; SABUNDJIAN, GAIANE
    During a nuclear power plant basic design accident, the containment integrity is a determining factor for the accident severity. The pressure and temperature conditions inside the containment in case of a Large Break Loss of Coolant Accident (LBLOCA) must be verified. This paper presents a containment pressure and temperature analysis methodology of a Brazilian PWR, Angra 2, using a code that simulates guillotine rupture - RELAP5 - and the COCOSYS code, which analyzes the containment pressure from the accident conditions. The Angra 2 containment behavior results during the design basis accidents studied - primary cooling system cold and hot legs guillotine ruptures - were satisfactory when compared to those presented in the Final Safety Analysis Report (FSAR / A2) and the pressure distributions were below the containment design pressure value (6.3bar).
  • 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.
  • Artigo IPEN-doc 25067
    Simulação de um SBLOCA em Angra 2 com o RELAP5
    2018 - BORGES, EDUARDO M.; SABUNDJIAN, GAIANE; CONTI, THADEU das N.; BRAZ FILHO, FRANCISCO A.; GUIMARAES, LAMARTINE N.F.
    O objetivo deste trabalho foi simular e avaliar o acidente básico de projeto de perda de refrigerante por pequena ruptura de 50 cm2 na perna fria do circuito primário da usina nuclear Angra 2. Nesta simulação, verificou-se que as lógicas de atuação do Sistema de Proteção do Reator (SPR) e do Sistema de Resfriamento de Emergência do Núcleo (SREN) de Angra 2 utilizadas nesta simulação funcionaram corretamente, mantendo a integridade do núcleo com temperaturas aceitáveis durante todo o evento. Os resultados obtidos foram satisfatórios quando comparados com os apresentados pelo Relatório Final de Análise de Segurança de Angra 2 (FSAR/A2).