WALMIR MAXIMO TORRES
48 resultados
Resultados de Busca
Agora exibindo 1 - 10 de 48
Artigo IPEN-doc 26385 Preliminary numerical analysis of the flow distribution in the core of a research reactor2019 - SCURO, NIKOLAS L.; ANGELO, GABRIEL; ANGELO, E.; TORRES, WALMIR M.; UMBEHAUN, PEDRO E.; ANDRADE, DELVONEI A. deThe 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.Artigo IPEN-doc 26349 Lower plenum holes for research reactor core flooding2019 - MAPRELIAN, EDUARDO; BELCHIOR JUNIOR, ANTONIO; TORRES, WALMIR M.Modern and high power pool type research reactors generally operate with upward flow in the core. They have a chimney above the core, where the heated fluid is suctioned by the pumps. It passes through the decay tank and is sent to the heat exchangers for the cooling and returns to the core. The pipes inside the reactor pool have passive valves (natural circulation valves) that allow the establishment of natural circulation between the core and the pool for the decay heat removal, when the pumps are inoperative. These valves also have the siphon-breaker function in case of Loss of Coolant Accidents (LOCA), avoiding the pool emptying. In some reactors, these valves are located above the core chimney to facilitate the maintenance. When a LOCA causes a water level below these valves, they loose the natural circulation function. If the water level is the same of the chimney top, the available fluid for the core cooling is only that contained in the chimney and core, and a significant quantity of water in the pool is unavailable for core cooling. To bypass this problem during the reactor design phase, the inclusion of small holes of 10 mm of diameter on the lower plenum lateral side is proposed. These holes will allow a flow path between the pool and the core. Theoretical calculations were performed and analyzed for different drilling configurations: 4, 6 8, and 10 holes. A theoretical analysis of the estimated leakage rate during normal operation and evaporation and replacement rates during a hypothetical LOCA were performed. The calculation results showed that the four configurations analyzed are able to supply the water evaporated from chimney. An experiment is being proposed to validate the theoretical calculations and the considered hypotheses.Artigo IPEN-doc 26346 Status of the development of a fuel assembly decay heat calorimeter for the IEA-R1 nuclear research reactor2019 - 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.Artigo IPEN-doc 26344 RMB experimental program on the hydrodynamical behavior of fuel assemblies2019 - TORRES, WALMIR M.; UMBEHAUN, PEDRO E.; MATTAR NETO, MIGUEL; BELCHIOR JUNIOR, ANTONIO; FREITAS, ROBERTO L.The Brazilian Multipurpose Reactor - RMB is a 30 MW pool type research reactor, that uses Materials Testing Reactor - MTR type fuel assemblies. It has a 5x5 square array core with 23 fuel assemblies and two in-core irradiation positions, operating with upward flow and average velocities nearly 10 m/s in the fuel plates channels. The IEA-R1 is a 5 MW pool type research reactor, which also has a 5x5 square array core with 19 standard fuel assemblies, four control fuel assemblies and a central beryllium irradiation device. It operates with downward flow nearly 1.8 m/s in the channels. In order to verify and provide data and information about the dynamical behavior of fuel assemblies under nominal and critical conditions, the experimental circuit ORQUÍDEA is being designed. This information will be very important for the licensing process of the fuel assembly before its use in the reactor core. This circuit will permits upward and downward flow and dynamical behavior of the fuel assemblies and its parts will be tested and verified. Flow rate, temperature, pressure and differential pressure transducers are the instruments of the circuit. Endurance and critical flow velocity tests will be performed. Dummy fuel assemblies will be used in the tests. It will be instrumented with pressure, strain-gages and flow velocity instruments. The COLIBRI experimental circuit is being designed to make tests that allow the studies of the fluid-structure phenomenology of fuel plates similar to those of the RMB fuel assemblies when subjected to high flow velocities, which can induce pressure differences between the channels formed by the fuel plates. Preliminary structural response studies of the plate’s behavior were performed using a Finite Element Analysis model generated by ANSYS Mechanical. The pressure loadings caused by the fluid flow were calculated using a Computational Fluid Dynamics model created with ANSYS CFX. The fluid-structure interactions will be verified for different channel configurations. In this circuit, vibrations and collapse of the dummy fuel plates will be tested. Experimental data will be compared with CFD (Computational Fluid Dynamics) calculations. This work presents a preliminary design for the ORQUÍDEA and COLIBRI experimental circuits to be built at the Instituto de Pesquisas Energéticas e Nucleares - IPEN of the Comissão Nacional de Energia Nuclear - CNEN.Artigo IPEN-doc 25814 Procedures for manufacturing an instrumented nuclear fuel element2019 - 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.Artigo IPEN-doc 25071 Two-phase flow bubble detection method applied to natural circulation system using fuzzy image processing2018 - BUENO, R.C.; MASOTTI, P.H.F.; JUSTO, J.F.; ANDRADE, D.A.; ROCHA, M.S.; TORRES, W.M.; MESQUITA, R.N. deNatural circulation cooling systems are currently used in new nuclear reactors. Over the last decades, research in these systems has focused in the study of flow and heat transfer parameters. A particular area of interest is the estimation of two-phase flow parameters by image processing and pattern recognition using intelligent processing. Several methods have been proposed to identify objects of interest in bubbly two-phase images. Edge detection is an important task to estimate flow parameters, in which the bubbles are segmented to obtain several features, such as void fraction, area, and diameter. However, current methods face difficulties in determining those parameters in high bubble-density two-phase flow images. Here, a new edge detection method is proposed to segment bubbles in natural circulation instability images. The new method (Fuzzy Contrast Standard Deviation – FUZCON) uses Fuzzy Logic and image standard deviation estimates of locally measured contrast levels. Images were obtained through an experimental circuit made of glass, which enables imaging flow patterns of natural circulation cycles at ambient pressure. The results indicated important improvements on edge detection efficiency for high void fraction estimation on high-density two-phase flow bubble images, when compared to classical detectors, without the need to use smoothing algorithms or human intervention.Relatório IPEN-doc 24918 Avaliação técnica amostral da instrumentação do LOOP 70 para utilização no Circuito Orquídea2018 - SANTOS, SAMUEL C.; SANTOS, SERGIO O. dos; TORRES, WALMIR M.Este relatório apresenta o resultado da avaliação técnica de uma amostragem de transmissores de pressão, transmissores de pressão diferencial e transmissores temperatura do LOOP 70 para verificação da viabilidade de utilização destes instrumentos no circuito experimental Orquídea.Relatório IPEN-doc 24917 Memorial de cálculo do Circuito Hidrodinâmico para Testes de Elementos Combustíveis (Orquídea)2018 - TORRES, WALMIR M.Este relatório apresenta o memorial de cálculo do circuito Orquídea, que é o circuito experimental que está sendo projetado para a realização de testes hidrodinâmicos em elementos combustíveis (ECs) com placas planas paralelas do tipo MTR, para reatores nucleares de pesquisa, e que deverá ser construído no IPEN. Esse circuito experimental será utilizado para a realização dos testes necessários para a qualificação dos elementos combustíveis do reator RMB. O dimensionamento do circuito considera as condições mais críticas de operação, ou seja, as condições necessárias para atingir a velocidade de escoamento crítica em um EC tipo MTR, com escoamento ascendente, como o do reator RMB.Artigo IPEN-doc 24804 Thermal hydraulic analysis improvement for the IEA-R1 research reactor and fuel assembly design modification2018 - 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. deThis 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.Artigo IPEN-doc 24758 Classification of natural circulation two-phase flow image patterns based on self-organizing maps of full frame DCT coefficients2018 - 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.