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 107
  • Artigo IPEN-doc 26920
    Levantamento bibliográfico sobre metodologias para elaboração de um banco de dados da saúde da população em casos de ocorrências de câncer
    2015 - CAVINATO, C.C.; ANDRADE, D.A.; DIZ, M.D.P.E.; SABUNDJIAN, G.
    As fontes alternativas de energia, incluindo a nuclear, apresentam vantagens com relação às externalidades, as quais podem ser identificadas e relacionadas com o termo, custo ambiental. Este termo, por sua vez, é uma externalidade negativa, que de alguma forma prejudica o meio ambiente e é convertida em termos econômicos, para então poder ser comparada aos demais custos de uma ação e/ou empreendimento. A fim de efetuar os cálculos em questão, são utilizados alguns softwares específicos, os quais possibilitam a conversão dos danos em termos econômicos e a inclusão do custo ambiental na análise de custo de determinado projeto. Uma das dificuldades encontradas na utilização destes softwares tem sido com relação a alguns dados de entrada, como por exemplo, os relativos à saúde da população em torno da instalação nuclear, que são muito deficientes. O objetivo deste trabalho é fazer um levantamento teórico correspondente às metodologias existentes e utilizadas para a elaboração de banco de dados de saúde pública no Brasil. A partir das metodologias encontradas para a formação deste tipo de banco de dados, posteriormente será desenvolvida uma metodologia focando na saúde (câncer fatal e não fatal) da população circunvizinha a uma instalação nuclear, para fins de cálculo do custo ambiental da mesma. Essa será aplicada ao público interno do Instituto de Pesquisas Energéticas e Nucleares (IPEN), como um pré-teste, para aquisição das informações de saúde desejadas.
  • Artigo IPEN-doc 26677
    Pixel-position-based lossless image compression algorithm
    2019 - CABRAL, EDUARDO L.L.; SABUNDJIAN, GAIANE; CONTI, THADEU das N.
    In this paper we present a novel lossless image compression method that is very simple and fast. The method uses linear prediction followed by arithmetic coding. Different prediction functions are used to estimate the intensity of image pixels. Two variants of the prediction algorithm are presented. One variant uses two different prediction functions and the other uses three different prediction functions. The position of the pixel in the image determines which prediction function is used. The method can be applied for images of any size and of high bit-depths. Standard images available in the literature are used to test the method. The compression ratios obtained with the proposed method are compared with the compression ratios obtained with the JPEG-LS and JPEG2000 methods and the results are satisfactory.
  • 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 26373
    MELCOR simulation of a severe accident scenario derived from a small break loca in a typical PWR with passive autocatalytic recombiners
    2019 - LEE, SEUNG M.; LAPA, NELBIA S.; SABUNDJIAN, GAIANE
    This work presents the simulation of a severe accident scenario in a referential model of pressurized water reactor, which came about from a rupture of 20cm2 in a cold leg of a reactor cooling system. The simulation was carried out on the MELCOR code using a model elaborated by the Global Research for Safety – Germany, with the passive autocatalytic recombiners implemented in almost every compartment in the containment. The efficacy and effectiveness of this well-known mitigating measure of severe accident management are demonstrated by means of a comparison with the case previously simulated without this measure using the same model. This referential reactor is important and very useful for the independent analysis of severe accidents in the Brazilian Angra 2 nuclear power plant in virtue of the similarity between both of them, so that after some proper modifications on this referential reactor’s model, it could be applied for the study of severe accidents in the other. In this sense, the result presented in this work is to be taken as an important reference for the severe accident analysis of Angra 2.
  • Artigo IPEN-doc 26353
    Virtual Reality tools for goods, food and beverage irradiation at IPEN's facilities as a nuclear technology teaching motivation
    2019 - PALADINO, PATRICIA A.; SABUNDJIAN, GAIANE; CABRAL, EDUARDO L.L.; JULIÃO, ARTHUR P.
    In this research a full-fledged and complete Virtual Reality (VR) environment will be wholly developed and then deployed as a kind of innovative means of widespread divulgation of one topic of nuclear science and nuclear technology most interesting application and its teaching; viz, that related to goods, beverages and mainly food irradiation practices, simulating a virtually guided visit to some of IPEN’s facilities and its already installed and operational scientific equipment, namely, the GAMMACELL irradiator, firstly targeting undergraduate and last year high school students and then, later, the interested general public. In this way, several programs and whole VR platforms, such as Unity, are used as powerful, professional tools for games and videogames development and it is expected that the final product will be made available packaged as an instructive videogame to the community of committed and interested users. Therefore, in doing so, some contemporary reasoned and still debated pedagogical recommendations will be handled and met, hopefully increasing students’ curiosity and good aptitudes towards the disseminated use of nuclear technologies nowadays. It is hoped that perhaps a modest contribution against the many undeserved prejudices and odd misconceptions still remaining nowadays regarding nuclear science development, results and applications, will be abated.
  • Artigo IPEN-doc 26352
    Virtual visit to nuclear research reactor IEA-R1
    2019 - SILVA, LEANDRO G.M. e; SABUNDJIAN, GAIANE
    The aim of this paper is to provide students, educators, and the general public with a virtual tool for learning about the peaceful use of nuclear technology and its importance to humanity. Using new technologies available in the market such as smartphones, software for the development of electronic games, virtual reality glasses, among others, we will virtually reproduce the facilities of the IEA-R1 nuclear research reactor, allowing anyone to perform a virtual and interactive visit to these facilities in a safe and didactic way. The use of virtual reality glasses and applications has been shown to be adequate in relation to the objectives proposed here.
  • 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.