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 25
  • 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 27183
    Total and partial loss of coolant experiments in an instrumented fuel assembly of IEA-R1 research reactor
    2020 - MAPRELIAN, EDUARDO; TORRES, WALMIR M.; BELCHIOR JUNIOR, ANTONIO; UMBEHAUN, PEDRO E.; BERRETTA, JOSE R.; SABUNDJIAN, GAIANE
    The safety of nuclear facilities has been a growing global concern, mainly after the Fukushima nuclear accident. Studies on nuclear research reactor accidents such as the Loss of Coolant Accident (LOCA), many times considered a design basis accident, are important for ensure the integrity of the plant. A LOCA may lead to the partial or complete uncovering of the fuel assemblies and it is necessary to assure the decay heat removal as a safety condition. This work aimed to perform, in a safe way, partial and complete uncovering experiments for an Instrumented Fuel Assembly (IFA), in order to measure and compare the actual fuel temperatures behavior for LOCA in similar conditions to research reactors. A test section for experimental simulation of Loss of Coolant Accident named STAR was designed and built. The IFA was irradiated in the IEA-R1 core and positioned in the STAR, which was totally immersed in the reactor pool. Thermocouples were installed in the IFA to measure the clad and fluid temperatures in several axial and radial positions. Experiments were carried out for five levels of uncovering of IFA, being one complete uncovering and four partial uncovering, in two different conditions of decay heat. It was observed that the cases of complete uncovering of the IFA were the most critical ones, that is, those cases presented higher clad temperatures when compared with partial uncovering cases, for the specific conditions of heat decay intensity and dissipation analyzed. The maximum temperatures reached in all experiments were quite below the fuel blister temperature, which is around 500 °C. The STAR has proven to be a safe and reliable experimental apparatus for conducting loss of coolant experiments.
  • Resumo IPEN-doc 24614
    A CFD numerical model for the flow distribution in a MTR fuel element
    2017 - ANDRADE, D.A.; ANGELOA, G.; ANGELO, E.; SANTOS, P.H.G.; OLIVEIRA, F.B.V.; TORRES, W.M.; UMBEHAUN, P.E.; SOUZA, J.A.B.; BELCHIOR JUNIOR, A.; SABUNDJIAN, G.; PRADO, A.C.
    Previously, an instrumented dummy fuel element (DMPV-01), with the same geometric characteristics of a MTR fuel element, was designed and constructed for pressure drop and flow distribution measurement experiments at the IEA-R1 reactor core. This dummy element was also used to measure the flow distribution among the rectangular flow channels formed by element fuel plates. A CFD numerical model was developed to complement the studies. This work presents the proposed CFD model as well as a comparison between numerical and experimental results of flow rate distribution among the internal flow channels. Numerical results show that the model reproduces the experiments very well and can be used for the studies as a more convenient and complementary tool.
  • Resumo IPEN-doc 24578
    Experiments of loss of coolant in the IEA-R1 reactor
    2017 - MAPRELIAN, E.; TORRES, W.M.; BELCHIOR JUNIOR, A.; UMBEHAUN, P.E.; SANTOS, S.C.; FRANÇA, R.L.; PRADO, A.C.; MACEDO, L.A.; SILVA, A.T. E; BERRETTA, J.R.; SABUNDJIAN, G.
    The Loss of Coolant Accident (LOCA) has been considered Design Basis Accident (DBA) for several kind of reactors. The test section for experimental (STAR) for simulation of LOCA, using the Instrumented Fuel Assembly (IFA) EC-208 was designed, assembled, commissioned, and used for the experiments at the IEA-R1 Reactor. The experiments were performed for five different levels of fuel uncovering and two heat decay conditions. The five levels consisted of one total and four partial uncovering of the IFA. The results obtained for each experiment were the section level and 13 IFA temperatures. A data acquisition system was used to record the process parameters. The STAR section has proved to be a very safe and efficient tool for fuel uncovering experiments to obtain thermal-hydraulic data for research and development, and for the data to be compared with safety analysis code calculations.
  • Resumo IPEN-doc 23885
    Heat transfer mode in the core of the Angra 2 nuclear power plant during small break loca obtained with RELAP5 code
    2013 - SABUNDJIAN, GAIANE; BELCHIOR JUNIOR, ANTONIO; CONTI, THADEU das N.; ROCHA, MARCELO da S.; ANDRADE, DELVONEI A. de; TORRES, WALMIR M.; MACEDO, LUIZ A.; UMBEHAUN, PEDRO E.; MESQUITA, ROBERTO N. de; LIMA, ANA C. de S.
    This work aims to identify the heat transfer mode with RELAP5/MOD3.2.gama code in the core of Angra 2 facility. The postulate accident is the Loss of Coolant Accident (LOCA) in the primary circuit for Small Break (SB), which is described in Chapter 15 of the Final Safety Analysis Report of Angra 2 (FSAR). The accident consists basically of the total break of the cold leg of Angra 2 facility. The rupture area considered was 380 cm2, which represents 100% of the primary circuit pipe °ow area. The Emergency Core Cooling System (ECCS) e±ciency is also tested in this accident. In this simulation, failure and repair criteria are adopted for the ECCS components in order to verify the system operation e±ciency - preserving the integrity of the reactor core and guaranteeing its cooling - as expected by the project design. SBLOCA accidents are characterized by a fast blowdown in the primary circuit to values that activate the low pressure injection system followed by the water injection from the accumulators. The thermal-hydraulic processes inherent to the accident phenomenon, such as hot leg vaporization and consequently core vaporization cause inappropriate °ow distribution in the reactor core that can lead to reduction in the core liquid level, up to the point that the ECCS is able to re°ood it. This work shows the mode numbers and the wall convection heat transfer used in the RELAP5 code that were accessed during the execution of the program. The results showed that the numerical simulations with RELAP5 were satisfactory and that the ECCS worked e±ciently, guaranteeing the integrity of the reactor core.
  • Artigo IPEN-doc 21137
    Characterization of physical properties of Alsub(2)Osub(3) and ZrOsub(2) nanofluids for heat transfer applications
    2015 - ROCHA, MARCELO S.; CABRAL, EDUARDO L.L.; SABUNDJIAN, GAIANE; YORIYAZ, HELIO; LIMA, ANA C.S.; BELCHIOR JUNIOR, ANTONIO; PRADO, ADELK C.; MADI FILHO, TUFIC; ANDRADE, DELVONEI A.; SHORTO, JULIAN M.B.; MESQUITA, ROBERTO N.; OTUBO, LARISSA; BAPTISTA FILHO, BENEDITO D.; PINHO, PRISCILA G.M.; RIBATSKY, GHERHARDT; MORAES, ANDERSON A.U. de
  • Artigo IPEN-doc 20980
    A CFD numerical model for the flow distribution in a MTR fuel element
    2015 - ANDRADE, DELVONEI A. de; ANGELO, GABRIEL; ANGELO, EDVALDO; SANTOS, PEDRO H. di G.; OLIVEIRA, FABIO B.V. de; TORRES, WALMIR M.; EMBEHAUN, PEDRO E.; SOUZA, JOSE A.B. de; BELCHIOR JUNIOR, ANTONIO; SABUNDJIAN, GAIANE; PRADO, ADELK de C.
  • Artigo IPEN-doc 20797
    Thermophysical characterization of Alsub(2)Osub(3) and ZrOsub(2) nanofluids as emergency cooling fluids of future generations of nuclear reactors
    2015 - ROCHA, MARCELO S.; CABRAL, EDUARDO L.L.; SABUNDJIAN, GAIANE; YORIYAZ, HELIO; LIMA, ANA C.S.; BELCHIOR JUNIOR, ANTONIO; PRADO, ADELK C.; MADI FILHO, TUFIC; ANDRADE, DELVONEI A.; SHORTO, JULIAN M.B.; MESQUITA, ROBERTO N.; OTUBO, LARISSA; BAPTISTA FILHO, BENEDITO D.; RIBATSKY, GHERHARDT; MORAES, ANDERSON A.U. de
  • Artigo IPEN-doc 18514
    ANGRA 2 samll break loca flow regime identification through RELAP5 code
    2012 - ROCHA, MARCELO da S.; SABUNDJIAN, GAIANE; BELCHIOR JUNIOR, ANTONIO; ANDRADE, DELVONEI A. de; TORRES, WALMIR M.; CONTI, THADEU das N.; MACEDO, LUIZ A.; UMBEHAUN, PEDRO N.; MESQUITA, ROBERTO N. de; MASOTTI, PAULO H.F.
  • Artigo IPEN-doc 10640
    Simulacao e analise do fenomeno de circulacao natural monofasica e bifasica no circuito experimental instalado na engenharia quimica POLI-USP, com o codigo RELAP5
    2005 - ANDRADE, D.A.; SABUNDJIAN, G.; UMBEHAUN, P.E.; TORRES, W.M.; BELCHIOR JUNIOR, A.; ROCHA, R.T.V.; FERNANDES, T.D.J.; CARVALHO, A.D.