PEDRO ERNESTO UMBEHAUN

PEDRO ERNESTO UMBEHAUN

(Fonte: Lattes)

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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)

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Preliminary numerical analysis of the flow distribution in the core of a research reactor
2019 - SCURO, NIKOLAS L.; ANGELO, GABRIEL; ANGELO, E.; TORRES, WALMIR M.; UMBEHAUN, PEDRO E.; ANDRADE, DELVONEI A. de
The thermal-hydraulic safety analysis of research reactors establishes the safety criteria to ensure the integrity of the fuel elements in the reactor core. It assures that all core components are being adequately cooled during operation. It is necessary to know if the average mass flow rate (and their standard deviation) among the fuel assemblies are enough to cool the power generated during operation. Once satisfied such condition, it allows the calculation of the maximum heat flux transferred from fuel assemblies to the coolant, and if the maximum cladding temperatures are below the limits set by the safety criteria. Among the objectives, this study presents a methodology for a preliminary three-dimensional numerical analysis of the flow distribution in the core of the IEA-R1 research reactor, under steady state condition. For this, the ANSYS-CFX® commercial code was used to analyze the flow dynamics in the core, and to visualize the velocity field. It was possible to conclude that a homogeneous flow distribution for all standard fuel assemblies were found, with 2.7% deviation from the average mass flow. What turned out to be negligible and can be assumed that there is a homogeneous distribution in the core. Complex structures were find in the computational domain. Once known the core flow dynamics, it allows future studies to determine whether the heat flux and temperature conditions abbeys thermal-hydraulic safety criteria.
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.
A CFD numerical model for the flow distribution in a MTR fuel element
2017 - ANDRADE, D.A.; ANGELOA, G.; ANGELO, E.; SANTOS, P.H.G.; OLIVEIRA, F.B.V.; TORRES, W.M.; UMBEHAUN, P.E.; SOUZA, J.A.B.; BELCHIOR JUNIOR, A.; SABUNDJIAN, G.; PRADO, A.C.
Previously, an instrumented dummy fuel element (DMPV-01), with the same geometric characteristics of a MTR fuel element, was designed and constructed for pressure drop and flow distribution measurement experiments at the IEA-R1 reactor core. This dummy element was also used to measure the flow distribution among the rectangular flow channels formed by element fuel plates. A CFD numerical model was developed to complement the studies. This work presents the proposed CFD model as well as a comparison between numerical and experimental results of flow rate distribution among the internal flow channels. Numerical results show that the model reproduces the experiments very well and can be used for the studies as a more convenient and complementary tool.
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.
Simulacao de acidentes tipo LOCA em Angra 2 com o codigo RELAP5/MOD3.2
2005 - ANDRADE, D.A.; SABUNDJIAN, G.; UMBEHAUN, P.E.; TORRES, W.M.
Analise termo-hidraulica das placas externas de um elemento combustivel tipo placa utilizado no reator de pesquisa IEA-R1
2004 - UMBEHAUN, P.E.; TORRES, W.M.; ANDRADE, D.A.
Dimensionamento termo-hidraulico de um dispositivo de irradiacao de mini-placas combustiveis para o reator IEA-R1
2002 - UMBEHAUN, P.E.; ANDRADE, D.A.; TORRES, W.M.; YAMAGUCHI, M.
Thermal hydraulic phenomenology for the heating process in a natural circulation facility
2009 - TORRES, WALMIR M.; MACEDO, LUIZ A.; MESQUITA, ROBERTO N.; MASOTTI, PAULO H.F.; LIBARDI, ROSANI M.P.; SABUNDJIAN, GAIANE; ANDRADE, DELVONEI A.; UMBEHAUN, PEDRO E.; CONTI, THADEU N.; FILHO, MAURO F.S.; MELO, GABRIEL R.
Comparison between experimental data and numerical modeling for the natural circulation phenomenon
2009 - SABUNDJIAN, GAIANE; ANDRADE, DELVONEI A.; UMBEHAUN, PEDRO E.; TORRES, WALMIR M.; MACEDO, LUIZ A.; CONTI, THADEU N.; MESQUITA, ROBERTO N.; MASOTTI, PAULO H.F.; PENHA, ROSANI M.L.; SILVA FILHO, MAURO F.; MELO, GABRIEL R.; BRAZ FILHO, FRANCISCO A.; BORGES, EDUARDO M.
Thermal-hydraulic design of a fuel mini-plate irradiator for the IEA-R1 research reactor
2003 - TORRES, W.M.; UMBEHAUN, P.E.; ANDRADE, D.A.; YAMAGUCHI, M.; SOUZA, J.A.B.