ROBERTO NAVARRO DE MESQUITA

ROBERTO NAVARRO DE MESQUITA

(Fonte: Lattes)

Resumo

Possui graduação em Física pela Universidade Estadual de Campinas (1987), mestrado em Física pela Universidade Estadual de Campinas (1991) e doutorado em Engenharia Mecânica pela Universidade de São Paulo (2002). Atualmente é tecnologista pleno da Comissão Nacional de Energia Nuclear. Tem experiência na área de Ciência da Computação, com ênfase em Sistemas de Inteligência Artificial, atuando principalmente nos seguintes temas: inteligência artificial, diagnóstico de defeitos em tubos, correntes parasitas (ECT), reconhecimento de padrões em imagens. (Texto extraído do Currículo Lattes em 27 dez. 2021)

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Improving the RELAP5 code modeling of the siphon break effect in a pool type research reactor
2024 - BELCHIOR JUNIOR, ANTONIO; SOARES, HUMBERTO V.; FREITAS, ROBERTO L.; MESQUITA, ROBERTO N. de; ROCHA, MARCELO da S.; ANDRADE, DELVONEI A. de
The pool water of a research reactor is used for emergency cooling of the reactor core. Siphon breakers are installed in the lines of the Core Cooling System to stop the loss of water from the pool due to the siphon effect during an accident involving piping ruptures. Previous studies discuss the effectiveness of siphon breakers based on the air inlet area and question the ability of one-dimensional thermo-hydraulic codes to model the siphon break devices. By means of comparison with experimental results, this work demonstrates the ability of the RELAP5/MOD3.3 code to model the performance of the siphon breaker. There was satisfactory agreement between the numerical and experimental results, showing that, as the air intake areas of the siphons decrease, their effectiveness also decreases, resulting in greater drainage of the pool water. For smaller air intake areas, the RELAP5/MOD3.3 code showed conservative results, overestimating the reactor pool water losses.
CFD Simulation of isothermal upward two-phase flow in a vertical annulus using interfacial area transport equation
2023 - CERAVOLO, FLAVIO E.; ROCHA, MARCELO da S.; MESQUITA, ROBERTO N. de; ANDRADE, DELVONEI A. de
This work presents a numerical simulation of a vertical, upward, isothermal two-phase flow of air bubbles and water in an annular channel applying a Computational Fluid Dynamics (CFD) code. For this, the Two-Fluid model is applied considering interfacial force correlations, namely: drag, lift, wall lubrication, turbulent dispersion, and virtual mass. The turbulence k-ε model effects and the influence of One-group Interfacial Area Transport Equation (IATE) are taken into account, in this case, the influence of two source term correlations for the bubble breakup and coalescence IATE is analysed. The work assesses whether the code properly represents the physical phenomenon by comparing the simulation results with experimental data obtained from the literature. Six flow conditions are evaluated based on two superficial liquid velocities and three void fractions in the bubbly flow regimen. The annular channel adopted has an outer pipe with an internal diameter of 38.1 mm and an inner cylinder of 19.1 mm. To represent this geometry, a three-dimensional mesh was generated with 160,000 elements, after a mesh sensitivity study. The void fraction distribution, taken radially to the flow section, is the main parameter analysed as well as interfacial area concentration, interfacial gas velocity, and bubble sizes distribution. The CFD model implemented in this work demonstrates satisfactory agreement with the reference experimental data but indicates the need for further improvement in the phase interaction models.
Two-phase flow void fraction estimation based on bubble image segmentation using Randomized Hough Transform with Neural Network (RHTN)
2020 - SERRA, PEDRO L.S.; MASOTTI, PAULO H.F.; ROCHA, MARCELO S.; ANDRADE, DELVONEI A. de; TORRES, WALMIR M.; MESQUITA, ROBERTO N. de
The International Atomic Energy Agency (IAEA) has been encouraging the use of passive cooling systems in new designs of nuclear power plants. Next nuclear reactor generations are intended to have simpler and robust safety resources. Natural Circulation based systems hold an undoubtedly prominent position among these. The study of limiting conditions of these systems has led to instability behavior analysis where many different two-phase flow patterns are present. Void fraction is a key parameter in thermal transfer analysis of these flow instability conditions. This work presents a new method to estimate void fraction from images captured of an experimental two-phase flow circuit. The method integrates a set of Artificial Neural Networks with a modified Randomized Hough Transform to make multiple scans over acquired images, using crescent-sized masks. This method was called Randomized Hough Transform with Neural Network (RHTN). Each different mask size is chosen according with bubble sizes, which are the main ‘objects of interest’ in this image analysis. Images are segmented using fuzzy inference with different parameters adjusted based on acquisition focus. Void fraction calculation considers the volume of the imaged geometrical section of flow inside cylindrical glass tubes considering the acquisition depth-of-field used. The bubble volume is estimated based on geometrical parameters inferred for each detected bubble. The image database is obtained from experiments performed on a vertical two-phase flow circuit made of cylindrical glass where flow-patterns visualization is possible. The results have shown that the estimation method had good agreement with increasing void fraction experimental values. RHTN has been very efficient as bubble detector with very low ‘false-positive’ cases (< 0.004%) due robustness obtained through integration between Artificial Neural Networks with Randomized Hough Transforms.
Classification of natural circulation two-phase flow image patterns based on self-organizing maps of full frame DCT coefficients
2018 - MESQUITA, ROBERTO N. de; CASTRO, LEONARDO F.; TORRES, WALMIR M.; ROCHA, MARCELO da S.; UMBEHAUN, PEDRO E.; ANDRADE, DELVONEI A.; SABUNDJIAN, GAIANE; MASOTTI, PAULO H.F.
Many of the recent nuclear power plant projects use natural circulation as heat removal mechanism. The accuracy of heat transfer parameters estimation has been improved through models that require precise prediction of two-phase flow pattern transitions. Image patterns of natural circulation instabilities were used to construct an automated classification system based on Self-Organizing Maps (SOMs). The system is used to investigate the more appropriate image features to obtain classification success. An efficient automated classification system based on image features can enable better and faster experimental procedures on two-phase flow phenomena studies. A comparison with a previous fuzzy inference study was foreseen to obtain classification power improvements. In the present work, frequency domain image features were used to characterize three different natural circulation two-phase flow instability stages to serve as input to a SOM clustering algorithm. Full-Frame Discrete Cosine Transform (FFDCT) coefficients were obtained for 32 image samples for each instability stage and were organized as input database for SOM training. A systematic training/test methodology was used to verify the classification method. Image database was obtained from two-phase flow experiments performed on the Natural Circulation Facility (NCF) at Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN), Brazil. A mean right classification rate of 88.75% was obtained for SOMs trained with 50% of database. A mean right classificationrate of 93.98% was obtained for SOMs trained with 75% of data. These mean rates were obtained through 1000 different randomly sampled training data. FFDCT proved to be a very efficient and compact image feature to improve image-based classification systems. Fuzzy inference showed to be more flexible and able to adapt to simpler statistical features from only one image profile. FFDCT features resulted in more precise results when applied to a SOM neural network, though had to be applied to the full original grayscale matrix for all flow images to be classified.
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.
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
Self-organizing maps applied to two-phase flow on natural circulation loop studies
2015 - 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. de
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
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.
The behaviour of ANGRA 2 nuclear power plant core for a small break LOCA simulated with RELAP5 code
2012 - SABUNDJIAN, GAIANE; ANDRADE, DELVONEI A.; BELCHIOR JUNIOR, ANTONIO; ROCHA, MARCELO da S.; CONTI, THADEU das N.; TORRES, WALMIR M.; UMBEHAUN, PEDRO E.; MESQUITA, ROBERTO N.; MASOTTI, PAULO H.F.