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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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 analysis of the IEA-R1 research reactor – a comparison between ideal and actual conditions
2017 - UMBEHAUN, P.E.; TORRES, W.M.
Thermal-hydraulic analysis were performed for the IEA-R1 research reactor considering ideal, estimated and actual flow rate conditions through the fuel elements. The ideal conditions were obtained dividing the total primary flow rate among the fuel elements and the estimated conditions were calculated using the computer program FLOW. The actual flow rate conditions were experimentally measured using an instrumented dummy fuel element. The results show that the actual conditions are far from ideal and calculated ones due to the high bypass flow that deviates the active reactor core through the irradiation devices, gaps, couplings, etc..Thus, the safety margins are smaller for the actual flow conditions.
Instrumented fuel assembly
2017 - UMBEHAUN, P.E.; ANDRADE, D.A.; TORRES, W.M.; RICCI, W.
The flow rate in the channel between two fuel assemblies is very difficult to estimate or measured. This flow rate is very important to the cooling process of the external plates. This work presents the project and construction of an instrumented fuel assembly with the objectives of perform more accurate safety analysis for the IEA-R1 reactor; determine the actual cooling conditions (mainly in the outermost fuel plate) and validate computer codes used for thermalhydraulic and safety analysis of research reactors. Fourteen thermocouples were installed in this instrumented fuel assembly. Four in each lateral channel, one in the inlet nozzle and one in the outlet nozzle. There are three thermocouples in each channel to measure the clad temperature and one thermocouple to measure the fluid temperature. Three series of experiments, for three different core configuration were carried out with the instrumented fuel assembly. In two experiments a box was installed around the core to reduce the cross flow between the fuel assembly and measure the impact in the temperatures of external plates. The experimental results obtained with the instrumented fuel element are very consistent with the phenomenology involved. Given the amount of information generated and its utility in the design, improvement and qualification in construction, assembly and manufacturing of instrumented fuel, this project turned out to be an important landmark on the thermal-hydraulic study of research reactor cores. The proposed solutions could be useful for other research reactors.
A MTR fuel element flow distribution measurement preliminary results
2017 - TORRES, W.M.; UMBEHAUN, P.E.; ANDRADE, D.A.; SOUZA, J.A.B.
An instrumented dummy fuel element (DMPV-01) with the same geometric characteristics of a MTR fuel element was designed and constructed for 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. Two probes with two pressure taps were constructed and assembled inside the flow channels to measure pressure drop and the flow velocity was calculated using pressure drop equation for closed channels. This work presents the experimental procedure and results of flow distribution measurement among the flow channels. Results show that the flow rate in the peripheral channels is 10 to 15% lower than the average flow rate. It is important to know the flow rate in peripheral channels because of uncertainties in values of flow rate in the open channel formed by two adjacent fuel elements. These flow rates are responsible by the cooling of external fuel plates.
Experiments of loss of coolant in the IEA-R1 reactor
2017 - MAPRELIAN, E.; TORRES, W.M.; BELCHIOR JUNIOR, A.; UMBEHAUN, P.E.; SANTOS, S.C.; FRANÇA, R.L.; PRADO, A.C.; MACEDO, L.A.; SILVA, A.T. E; BERRETTA, J.R.; SABUNDJIAN, G.
The Loss of Coolant Accident (LOCA) has been considered Design Basis Accident (DBA) for several kind of reactors. The test section for experimental (STAR) for simulation of LOCA, using the Instrumented Fuel Assembly (IFA) EC-208 was designed, assembled, commissioned, and used for the experiments at the IEA-R1 Reactor. The experiments were performed for five different levels of fuel uncovering and two heat decay conditions. The five levels consisted of one total and four partial uncovering of the IFA. The results obtained for each experiment were the section level and 13 IFA temperatures. A data acquisition system was used to record the process parameters. The STAR section has proved to be a very safe and efficient tool for fuel uncovering experiments to obtain thermal-hydraulic data for research and development, and for the data to be compared with safety analysis code calculations.
Determination of pressure loss coefficients in the elements of the IEA-R1 reactor nuclei
2017 - CASTRO, A.J.A. de; UMBEHAUN, P.E.
The flow distribution in the different elements that compose the core of the IEAR1 reactor is one of the main parameters for its thermo-hydraulic analysis. Currently this distribution is estimated with the code "FLOW" that uses existing correlations in the literature for the estimation of the singular and distributed pressure losses. In order to validate the code, a test bench was set up to survey the load loss in the elements that make up the reactor core for different levels of flow in the elements.
Comissioning of the IEA-R1 nuclear reactor new heat exchanger
2017 - CASTRO, A.J.A. de; UMBEHAUN, P.E.; CARVALHO, M.R.
This work presents results on the commissioning of the new heat exchanger of the IEA-R1 nuclear reactor in the occasion of its operational power upgrade from 2 MW to 5 MW, in comparison to the values calculated in the project of IESA Design and Equipments Company. This reactor is a swimming pool type, light water moderated and with graphite reflectors, used for research purposes and medical radioisotopes production. During monitoring procedures, issues were observed on the reactor operation at 5 MW mainly due to the ageing of the reactor’s oldest heat exchanger (TC-A) and excessive vibrations at high flow rates on the other installed heat exchanger (TC-B). So it was decided to provide a new IESA heat exchanger with 5 MW capacity to definitely substitute the TC-A heat exchanger. The results show that the IEA-R1 nuclear reactor can be operated safely and continuously at 5 MW with the new IESA heat exchanger.
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.
Comissioning of the heat exchanger for the research nuclear reactor IEA-R1
2008 - CASTRO, ALFREDO J.A. de; CASSIANO, DOUGLAS A.; UMBEHAUN, PEDRO E.; CARVALHO, MARCOS R. de; FRAJNDLICH, ROBERTO