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.
Status of the development of a fuel assembly decay heat calorimeter for the IEA-R1 nuclear research reactor
2019 - PRADO, ADELK C.; ANDRADE, DELVONEI A.; UMBEHAUN, PEDRO E.; TORRES, WALMIR M.; BELCHIOR JUNIOR, ANTONIO; PENHA, ROSANI M.L.
The heat release due to decay of fission products following a nuclear reactor shutdown is important matter for determining cooling requirements as well as for predicting postulated accident consequences. Accurate evaluation of decay heat can also potentially provide independent data for the cross examination of fuel burnup calculations, which is useful where few resources are available for examination of spent fuel. The evaluation of decay heat from unloaded fuel assemblies of the IEA R1 research reactor was proposed in order to seize that opportunity. With that purpose a special measuring device is under development at the Nuclear and Energy Research Institute (IPEN). Since average heat flux as low as 0.1W/cm2 is expected and since decay heat release must be accurately evaluated, the device design had to overcome the difficulties of measuring small amounts of heat released over a large boundary surface. The design had also to ensure the safe cooling of the fuel assemblies and proper radiological protection for the personnel. In view of the tight constraints, a novel design was adopted. The device features a submersible measurement chamber, which allows all measurement procedures to be performed without removing the fuel assemblies from the reactor pool, and an array of semiconductor thermoelectric modules, which provides highly accurate decay power measurements. The assemblage of the device is currently in progress, the main parts have already been acquired or manufactured and key components passed partial tests. Commissioning and main experiments will be performed up to the end of 2019.
A CFD analysis of blockage length on a partially blocked fuel rod
2019 - SCURO, N.L.; UMBEHAUN, P.E.; ANGELO, E.; ANGELO, G.; ANDRADE, D.A.
After a loss of coolant accident (LOCA), fuel rods may balloon. The swelling can partially block the flow channel, affecting the coolability during reflood phase. In order to analyze the influence of blockage length, using a radial block-age of 90%, varying just the blockage length, many steady state numerical simulations has been done using Ansys-CFX code to verify thermal-hydraulic properties according to different forced cooled conditions. Temperature peaks are observed on cladding, followed by a temperature drop. A 5x5 fuel assembly, with 9 centered ballooned fuel rod, flow redistribution inside channels can also be captured, indicating an overheating zone. Therefore, this study conclude, for the same boundary conditions, the longer the blockage length originated after LOCA events, the higher are the clad temperatures, indicating the possibility of overheat during transient conditions on reflood.
Procedures for manufacturing an instrumented nuclear fuel element
2019 - DURAZZO, M.; UMBEHAUN, P.E.; TORRES, W.M.; SOUZA, J.A.B.; SILVA, D.G.; ANDRADE, D.A.
The IEA-R1 is an open pool research reactor that operated for many years at 2 MW. The reactor uses plate type fuel elements which are formed by assembling eighteen parallel fuel plates. During the years of reactor operation at 2 MW, thermohydraulic safety margins with respect to design limits were always very high. However, more intense oxidation on some external fuel plates was observed when the reactor power was increased to 5 MW. At this new power level, the safety margins are significantly reduced due to the increase of the heat flux on the plates. In order to measure, experimentally, the fuel plate temperature under operation, an instrumented fuel element was constructed to obtain temperature experimental data at various positions of one or more fuel plates in the fuel element. The manufacturing method is characterized by keeping the original fuel element design specifications. Type K stainless sheathed thermocouples are mounted into supports pads in unrestricted positions. During the fuel element assembling, the supports pads with the thermocouples are mechanically fixed by interference between two adjacent fuel plates. The thermocouple wires are directed through the space existing at the bottom of the mounting slot where the fuel plate is fixed to the side plates. The number of thermocouples installed is not restricted and depends only on adaptations that can be made on the mounting slots of the standard fuel element side plates. This work describes the manufacturing procedures for assembling such an instrumented fuel element.
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.
Transient cfd analysis of the flow inversion of the nuclear research reactor IEA-R1
2018 - SCURO, N.L.; SANTOS, P.G.; UMBEHAUN, P.E.; ANDRADE, D.A.; ANGELO, E.; ANGELO, G.
The IEA-R1 research reactor works with a downflow direction, but after pumps shutdown during a LOFA test, the reactor shutdown. The heat decay will be removed by natural convection, which is an upward flow, originating flow inversion. Using the Instrumented Fuel Element designed at the Institute for Energy and Nuclear Research (IPEN), the loss of flow accident (LOFA) was analyzed along instrumented fuel plates. The preliminary results showed temperature peaks during inversion, which is as much representative as in nominal operation at 3.5MW. Therefore, these experimental data lead a construction and validation of a transient three-dimensional numerical analysis for a single fuel channel using the ANSYS-CFX® commercial code. The numerical results show improvement in obtaining more properties, e.g., wall heat transfer coefficient, which is usually obtained through empirical correlations.
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.
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.
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.