GAIANE SABUNDJIAN
ORCID 
ScopusID 
ResearcherID 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)
Lattes 
Orientações Projetos de Pesquisa
Unidades Organizacionais
Cargo
31 resultados
Resultados de Busca
Agora exibindo 1 - 10 de 31
Artigo IPEN-doc 28268 RELAP5 code simulation of the Angra2 pressurizer surge line accident2021 - SABUNDJIAN, GAIANE; PACHECO, RAFAEL R.; CONTI, THADEU das N.; LIMA, ANA C. de S.; SANCHES, ANDREAArtigo 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 Plant2019 - 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 26373 MELCOR simulation of a severe accident scenario derived from a small break loca in a typical PWR with passive autocatalytic recombiners2019 - LEE, SEUNG M.; LAPA, NELBIA S.; SABUNDJIAN, GAIANEThis 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 24026 Identification of flow regimes and heat transfer modes in ANGRA2 core during the simulation of the small break loss of coolant accident of 250 cm2 in the cold leg of primary loop using RELAP5 code2017 - BORGES, EDUARDO M.; SABUNDJIAN, GAIANEThe aim of this paper is 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 250cm2 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 Angra2 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 24022 PCRELAP5 - a visual graphic preprocessor for RELAP52017 - MONACO, DANIEL F.; SABUNDJIAN, GAIANEThe aim of this work is to develop PCRELAP5, a visual preprocessor for RELAP5, reducing time, effort and maintenance costs spent in new projects for RELAP5. This preprocessor allows user to draw new nuclear power plant nodalization in a completely interactive way, and input parameters for each node in a more user-friendly experience. Once parameters are changed on screen, the input cards of RELAP5 code are changed in real time. RELAP5 users will have a tool to reduce time and effort for new studies and existing projects. Therefore, this project proposes to significantly leverage studies related to nuclear accident analysis, making the RELAP5 code more user-friendly. In order to demonstrate this preprocessor capability, the CANON experiment will be used as an example. The PCRELAP5 preprocessor is being developed using Microsoft® Visual Studio® as a Microsoft® Excel® add-in, due to the low cost of distribution and maintenance, and also allowing new RELAP5 projects be leveraged by the MS Excel® flexibility.Artigo IPEN-doc 23821 BEPU-FSAR: a new paradigm in nuclear reactor safety2017 - MENZEL, F.; SABUNDJIAN, G.; DAURIA, F.To perform an entire FSAR based on BEPU (Best Estimated Plus Uncertainty), a homogenization of the analysis is proposed. The first step towards BEPU-FSAR requires identification and characterization of the FSAR parts where the numerical analyses are needed. The next step consists of creating a list of key technological areas where the relations between so-called key disciplines and the key topics are established. Considering the successful applications of BEPU methodology to the Chapter 15 of FSAR performed in the last two decades (Atucha II NPP, Angra 1 and 2), one can conclude that this methodology is feasible, which encourage to extended its range of use to the other technological areas of FSAR (e.g. seismology, radioprotection, etc.), and therefore to demonstrate the industrial worth and interest. The future step of this work will mainly be focused on the propagation of this expertise into the remaining technical areas of FSAR.Artigo IPEN-doc 21115 Commissioning of the star test section for experimental simulation of loss coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor2015 - MAPRELIAN, EDUARDO; TORRES, WALMIR M.; PRADO, ADELK C.; UMBEHAUN, PEDRO E.; FRANÇA, RENATO L.; SANTOS, SAMUEL C.; MACEDO, LUIZ A.; SABUNDJIAN, GAIANEArtigo IPEN-doc 21073 Self-organizing maps applied to two-phase flow on natural circulation loop studies2015 - CASTRO, LEONARDO F.; CUNHA, KELLY de P.; ANDRADE, DELVONEI A. de; SABUNDJIAN, GAIANE; TORRES, WALMIR M.; MACEDO, LUIZ A.; ROCHA, MARCELO da S.; MASOTTI, PAULO H.F.; MESQUITA, ROBERTO N. deArtigo IPEN-doc 21060 Flow regimes and heat transfer modes identification in ANGRA 2 core, during small break in the primary loop with area of 100 cmsup(2), simulated with RELAP5 code2015 - BORGES, EDUARDO M.; SABUNDJIAN, GAIANEArtigo IPEN-doc 21033 Methodology of a PWR containment analysis during a thermal-hydraulic accident2015 - SILVA, DAYANE F.; SABUNDJIAN, GAIANE; LIMA, ANA C.S.