WALMIR MAXIMO TORRES
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Relatório IPEN-doc 29806 Relatório IPEN-doc 29805 Geração de seções de choque para o ECI-RMB2023 - YAMAGUCHI, MITSUO; TORRES, WALMIR M.Este relatório apresenta a metodologia usada para a geração das seções de choque do Elemento Combustível Instrumentado ECI-RMB, que será irradiado no núcleo do Reator IEA-R1 e posteriormente permitirá a realização de testes não destrutivos (espectrometria gama, medida de espessura e inspeção visual) das placas removíveis representativas do reator RMB. Para possibilitar a sua irradiação no núcleo do reator IEA-R1, o ECI-RMB está sendo fabricado considerando as dimensões externas do elemento combustível padrão (EC) do IEA-R1. O ECI-RMB possui 16 placas combustíveis, sendo 10 placas padrão do EC do IEA-R1 (fixas) e 6 placas representativas do EC do RMB (removíveis); e ainda uma placa central espessa (6 mm) de alumínio ocupando as posições das 2 (duas) placas centrais para permitir a passagem de um detector de nêutrons SPND (Self Powered Neutron Detector) e de 2 (dois) termopares para medida das temperaturas do fluido na entrada e saída. A densidade de urânio nas placas combustíveis do EC do IEA-R1 é de 3,0 g/cm3 em um cerne de 0,76 mm de espessura, enquanto que a densidade de urânio nas placas combustíveis do EC do RMB é de 3,7 g/cm3 em um cerne de 0,61 mm. As placas do EC padrão do IEA-R1 têm espessura de 1,52 mm, enquanto que as placas do EC padrão RMB têm 1,35 mm de espessura. Todas essas diferenças, com relação ao EC padrão do IEA-R1, mostram a necessidade da geração das seções de choque para o ECI-RMB visando verificar a sua influência quando estiver sendo irradiado no núcleo do IEA-R1, e principalmente para verificar se nenhum limite de segurança será violado. Este relatório apresenta, nas Tabelas 2 e 3, os parâmetros utilizados para a geração das seções de choque o ECI-RMB, as quais foram utilizadas para o cálculo da distribuição de potência nas placas combustíveis, a qual será utilizada na análise termo-hidráulica do núcleo. Para a determinação das densidades de potência nas placas, utilizam-se os programas LEOPARD para geração das seções de choque do combustível, HAMMER-TECHNION para geração das seções de choque da guia de alumínio do SPND e 2DB e CITATION para o cálculo do núcleo.Relatório IPEN-doc 29651 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 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 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 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.Resumo IPEN-doc 24614 A CFD numerical model for the flow distribution in a MTR fuel element2017 - 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.Capítulo IPEN-doc 21750 IEA-R1 Nuclear Reactor: Facility specification and experimental results2015 - UMBEHAUN, P.E.; ANDRADE, D.A. de; TORRES, W.M.; RICCI FILHO, W.Artigo IPEN-doc 21115 Commissioning of the star test section for experimental simulation of loss coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor2015 - MAPRELIAN, EDUARDO; TORRES, WALMIR M.; PRADO, ADELK C.; UMBEHAUN, PEDRO E.; FRANÇA, RENATO L.; SANTOS, SAMUEL C.; MACEDO, LUIZ A.; SABUNDJIAN, GAIANE