SEUNG MIN LEE
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Artigo IPEN-doc 28295 Whole body dose due to station blackout at Angra 2 nuclear power plant2021 - AGUIAR, A.S.; LEE, S.M.; SABUNDJIAN, G.Artigo IPEN-doc 28265 Radiation shielding for a nuclear fusion device with inertial electrostatic confinement2021 - LEE, S.M.; YORIYAZ, H.; CABRAL, E.L.L.Capítulo IPEN-doc 27999 Calculation of the dose for public individuals due to a severe accident at the Angra 2 nuclear plant, Brazil2021 - AGUIAR, ANDRE S. de; LEE, SEUNG M.; SABUNDJIAN, GAIANEThrough a severe accident at nuclear power plant Angra 2, the whole body dose effective of the individuals members of the public located in the Emergency Planning Zones (EPZs) will be calculated, and later, the protective actions in these EPZs will be analyzed. Two different scenarios of radionuclide release into the atmosphere will be considered. In the first scenario, 2 h of the release of Xe, Cs, Ba, and Te, and the second scenario, 168 h of release.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 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 26365 Development of neutron shielding for an inertial electrostatic confinement nuclear fusion device2019 - LEE, SEUNG M.; YORIYAZ, HELIO; CABRAL, EDUARDO L.L.This work aims to develop a suitable neutron shielding for an Inertial Electrostatic Confinement Nuclear Fusion device (IECF). Neutrons are generated in the IECF device as results of nuclear fusion reactions and their detection is fundamental for the development of the IECF device, because experimental data is needed to perform efficiency analysis and model validation. Nevertheless, it is essential to moderate the neutrons down to the thermal state to make it possible to detect those using conventional detectors. Therefore, to properly measure the fast neutron generation rate by the IECF device it is necessary to previously elaborate a detailed neutron transport model between the IECF device and the radiation shielding, where the neutron detector will be located. In this work, a model is elaborated using the Monte Carlo N-Particle Code and is used to design the required radiation shielding for the device. Later, the same model will be used to determine the proportionality factor between the fast neutron generation in the IECF device and the thermal neutron population in the shielding.Artigo IPEN-doc 25017 Simulation of a station black out at the Angra 2 NPP with Melcor Code2017 - LAPA, NELBIA; MARTINS, LUIZ C.; MADEIRA, ALZIRA; WELLELE, OLIIVER; SABUNDJIAN, GAIANE; LEE, SEUNG; STEINROTTER, THOMASThe interest in evaluating the level of resistance of a nuclear power plant in response to an accident that exceeds the project bases, increased significantly after the Fukushima-Daiichi accident. Melcor is an integrated code, developed by Sandia National Laboratories, used to model and simulate the evolution of severe accidents in nuclear power plants. The Melcor modeling is general and flexible, making use of a “control volume” approach in describing the thermal hydraulic response of the plant. Reactor-specific geometry is imposed only in modeling the reactor core. The reactor cooling circuit and the four SG are represent by two model-loops, a single loop with the pressurizer and an agglutinated triple loop. All active safety systems which depend on AC power are assumed to be unavailable in this analysis. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system. The passive severe accident management measures primary bleed, secondary side bleed, passive injection from feedwater system and firefighting pool available. In Brazil there is the Almirante Álvaro Alberto Nuclear Power Plant that has two plants in operation, and one of them is Angra 2, which started operating in 2001. This unit is a pressurized water reactor type with electrical output of about 1350 MW. The objective of this work is to present a summary of the severe accident caused by a station black out condition using the Melcor 1.8.6 code. The main result of the study is an evaluation of RPV lower head integrity during a severe accidents scenario. The results will be useful to an independent assessment into the detailed processes involved by the management guidelines for one scenario severe accident in Angra 2.Artigo IPEN-doc 24941 Support to the nuclear safety regulator of Brazil (CNEN) through an INSC project2017 - LAPA, N.S.; PEREIRA, L.C.M.; MADEIRA, A.A.; WELLELE, O.J.M.; SABUNDJIAN, G.; LEE, S.M.; ARO, I.; STEINROTTER, T.; VARELA, J.G.; VALKONEN, J.; PILJUGIN, G.R.S.; FURIERI, E.; RODRIGUEZ, J.The paper introduces the European Union funded cooperation between the Brazilian nuclear regulatory body (CNEN) and a consortium of several European organizations. The still ongoing cooperation started in 2011 and has provided CNEN with insights and complementary information for licensing and regulatory activities on different reactor nuclear safety issues. The support is described by examples relating to the Severe Accident Management Program (SAMP) of the Angra 2 nuclear power plant NPP and the safety of digital instrumentation and control systems (DI&C) of Angra 3. The goal of the support regarding the SAMP is the review of the SAMP - under the licensing process by CNEN – with the focus on the new procedures and equipment implemented after the Fukushima accident. In parallel, a MELCOR simulation model of Angra 2 has been developed to perform independent calculations in order to support the assessment of the safety analysis presented in the Angra 2 Severe Accident Management Guides (SAMG). The support regarding DI&C is focused on regulatory issues concerning the review and assessment of digital instrumentation and control systems (DI&C) of the Angra 3 NPP under the licensing process by CNEN – providing CNEN with insights and complementary information on licensing experiences of new reactors with a DI&C architecture and technology similar to that of the Angra 3 NPP.Artigo IPEN-doc 24105 Study on fatal and nonfatal cancer cases occured in different regions of São Paulo city2017 - COELHO, TALITA S.; DIAS, BRUNA T.M.; SABUNDJIAN, GAIANE; LEE, SEUNG M.; DIZ, MARIA D.P.E.; FERNANDES, MARCO A.R.Artigo IPEN-doc 24024 Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS line using melcor code2017 - LEE, SEUNG M.; SABUNDJIAN, GAIANEThe aim of this work was to simulate a severe accident at a typical PWR caused by break in Emergency Core Cooling System (ECCS) line of a hot leg using the MELCOR code. The nodalization of this typical PWR was elaborated by the Global Research for Safety (GRS) and provided to the CNEN for analysis of the severe accidents at the Angra 2, which is similar to that PWR. Although both of them are not identical the results obtained for that typical PWR may be valuable because of the lack of officially published calculation for Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes after the break in the hot leg were calculated as well as degree of core degradation and hydrogen concentration in containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management (SAMG) by adding associated conditions for each measure in its input.