PEDRO ERNESTO UMBEHAUN

PEDRO ERNESTO UMBEHAUN

(Fonte: Lattes)

ORCID

ScopusID

ResearcherID

Resumo

Possui graduação em Engenharia Mecânica pela Fundação Educacional Inaciana Padre Sabóia de Medeiros -FEI (1985), mestrado em Tecnologia Nuclear pela Universidade de São Paulo (2000) e doutorado em Tecnologia Nuclear pela Universidade de São Paulo (2016). Atualmente é Tecnologista Sênior no Instituto de Pesquisas Energéticas e Nucleares da Comissão Nacional de Energia Nuclear. Tem experiência na área de Engenharia Nuclear, com ênfase em Transferência de Calor, atuando principalmente nos seguintes temas: termo-hidráulica de núcleo de reatores nucleares, engenharia nuclear, reatores de pesquisa, e reator nuclear de potência. Atualmente professor convidado na Escola Politécnica da Universidade de São Paulo nas disciplinas Termohidráulica de Sistemas de Geração de Potência I e II. (Texto extraído do Currículo Lattes em 4 maio 2023)

Lattes

Página do item completo

Resultados de Busca

Agora exibindo 1 - 10 de 20

Critical velocity experimental assessment in flat plate fuel element for nuclear research reactor
2022 - ANDRADE, D.A.; MANTECON, J.G.; MESQUITA, R.N.; MATTAR NETO, M.; UMBEHAUN, P.E.; TORRES, W.M.
Aluminum-coated plates, containing a uranium silicide (U3Si2) meat dispersed in an aluminum matrix, are commonly used in the fuel elements of Material Testing Reactors (MTRs). These fuel elements are typically comprised of narrow channels formed by parallel flat plates, which allow coolant flow to remove the heat of fission reactions. It is important to mention that the thickness of the plates is much smaller than their width and height. The high flow rates needed to ensure efficient fuel-element cooling may cause fuel-plate mechanical failures due to instability induced by the flow in the channels. In the case of critical velocity, excessive permanent deflections of these plates can cause blockage of the flow channels and lead to overheating. An experimental facility that simulates a plate-like fuel element with three coolant channels was developed for this work. The test-section dimensions were based on the Fuel Element design of the Brazilian Multipurpose Reactor (RMB), project being coordinated by the National Commission of Nuclear Energy (CNEN). Experiments were performed to reach Miller's critical velocity condition. This critical condition was reached at 14.5 m/s leading to consequent plastic deformation of the fuel plates.
Thermal hydraulic analysis improvement for the IEA-R1 research reactor and fuel assembly design modification
2018 - UMBEHAUN, PEDRO E.; TORRES, WALMIR M.; SOUZA, JOSE A.B.; YAMAGUCHI, MITSUO; SILVA, ANTONIO T. e; MESQUITA, ROBERTO N. de; SCURO, NIKOLAS L.; ANDRADE, DELVONEI A. de
This paper presents the sequence of activities to improve the thermal hydraulic analysis of the IEA-R1 research reactor to operate in safe conditions after power upgrade from 2 to 5 MW and core size reduction from 30 to 24 fuel assemblies. A realistic analysis needs the knowledge of the actual operation conditions (heat flow, flow rates) beyond the geometric data and the uncertainties associated with manufacturing and measures. A dummy fuel assembly was designed and constructed to measure the actual flow rate through the core fuel assemblies and its pressure drop. First results showed that the flow distribution over the core is nearly uniform. Nevertheless, the values are below than the calculated ones and the core bypass flow rate is greater than those estimated previously. Based on this, several activities were performed to identify and reduce the bypass flow, such as reduction of the flow rate through the sample irradiators, closing some unnecessary secondary holes on the matrix plate, improvement in the primary flow rate system and better fit of the core components on the matrix plate. A sub-aquatic visual system was used as an important tool to detect some bypass flow path. After these modifications, the fuel assemblies flow rate increased about 13%. Additional tests using the dummy fuel assembly were carried out to measure the internal flow distribution among the rectangular channels. The results showed that the flow rate through the outer channels is 10% - 15% lower than the internal ones. The flow rate in the channel formed between two adjacent fuel assemblies is an estimated parameter and it is difficult to measure because this is an open channel. A new thermal hydraulic analysis of the outermost plates of the fuel assemblies takes into account all this information. Then, a fuel design modification was proposed with the reduction of 50% in the uranium quantity in the outermost fuel plates. In order to avoid the oxidation of the outermost plates by high temperature, low flow rate, a reduction of 50% in the uranium density in the same ones was shown to be adequate to solve the problem.
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.
Study of the natural circulation phenomenon for nuclear reactors
2010 - CONTI, THADEU das N.; SABUNDJIAN, GAIANE; UMBEHAUN, PEDRO E.; MESQUITA, ROBERTO N.
Study of the natural circulation phenomenon for nuclear reactors
2010 - CONTI, THADEU das N.; SABUNDJIAN, GAIANE; TORRES, WALMIR M.; MACEDO, LUIZ A.; ANDRADE, DELVONEI A.; UMBEHAUN, PEDRO E.; MESQUITA, ROBERTO N.
Analise teorico/experimental do fenomeno de circulacao natural
2010 - SABUNDJIAN, GAIANE; CONTI, THADEU N.; TORRES, WALMIR M.; MACEDO, LUIZ A.; ANDRADE, DELVONEI A.; UMBEHAUN, PEDRO E.; MESQUITA, ROBERTO N.; SILVA FILHO, MAURO F.; BRAZ FILHO, FRANCISCO A.; BORGES, EDUARDO M.
Thermal hydraulic phenomenology in a natural circulation circuit
2010 - TORRES, WALMIR M.; MACEDO, LUIZ A.; MESQUITA, ROBERTO N.; MASOTTI, PAULO H.F.; SABUNDJIAN, GAIANE; ANDRADE, DELVONEI A.; UMBEHAUN, PEDRO E.
Analise teorico e experimental do fenomeno de circulacao natural
2008 - SABUNDJIAN, GAIANE; ANDRADE, DELVONEI A. de; UMBEHAUN, PEDRO E.; TORRES, WALMIR M.; CASTRO, ALFREDO J.A. de; CONTI, THADEU das N.; MASOTTI, PAULO H.F.; MESQUITA, ROBERTO N. de; PALADINO, PATRICIA A.; BRAZ FILHO, FRANCISCO A.; BORGES, EDUARDO M.; BELCHIOR JUNIOR, ANTONIO; ROCHA, RICARDO T.V. da; DAMY, OSVALDO L.A.
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.