WALMIR MAXIMO TORRES

WALMIR MAXIMO TORRES

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Verification and validation of seven turbulence models for a natural circulation loop under transient conditions
2024 - ANGELO, G.; ANGELO, E.; SCURO, N.L.; TORRES, W.M.; ANDRADE, D.A.
A numerical study of the vertical heater, vertical cooler (VHVC) natural circulation loop (NCL) at IPEN/CNEN-SP was conducted using a three-dimensional and transient mathematical model analyzed with the commercial software ANSYS CFX. The study focused on the stable and single-phase flow regime, with a Rayleigh number ranging from zero to 2.8×108. Seven turbulence models have been benchmarked: Zero Equation, Eddy Viscosity Transport Equation (EVTE), k−ω, k−ɛ, Shear Stress Transport (SST), Reynolds Stress (SSG), and Detached Eddy Simulation (DES). The results of these models were compared against each other and against experimental results obtained specifically for this purpose, focusing on the spatial distribution and temporal evolution of temperature at various points in the natural circulation loop. Among all tested models, the k−ɛ model demonstrated superior performance with the lowest average deviation, exhibiting lower initial turbulence production and buoyancy effects than the more complex models. This behavior suggests that the k−ɛ model is more accurate in predicting temperature distribution and is a better choice for transient flow analysis in natural circulation loops with similar geometries to those presented in this study.
Assessment of the IEA-R1 nuclear reactor using a nonstandard fuel assembly with six fuel plates of the Brazilian Multipurpose Reactor
2024 - SOARES, HUMBERTO V.; TORRES, WALMIR M.; UMBEHAUN, PEDRO E.; BELCHIOR, ANTONIO; ANDRADE, DELVONEI A. de
In order to qualify the fuel plates of the Brazilian Multipurpose Reactor (RMB), a nonstandard Instrumented Fuel Assembly (IFA) was designed and is being constructed to be burned in the IEA-R1 nuclear research reactor. IFA has fuel plates of different uranium densities (10 fixed fuel plates of 3.0 gU/cm3 – IEA-R1 standard; 6 removable fuel plates of 3.7 gU/cm3 – RMB; and a central aluminum plate). This paper is the first step to demonstrate that IEA-R1 can safely operate with this IFA. To verify the IFA thermal behavior inside the IEA-R1 core during reactor operation and certify the no power peaks occurrence, the power distribution was calculated for each fuel plate. LEOPARD and HAMMER-TECHNION codes were utilized to calculate the core thermal neutron cross section and CITATION code to calculate the core power distribution. Calculations were performed for 5 MW reactor power considering the IFA placed in a core peripheral position. The RMB fuel plates average power was 4.73 % higher compared to IEA-R1 fuel plates. This was expected due to the higher density of uranium in these plates. The power of each IFA fuel plate was compared with a fresh IEA-R1 Fuel Assembly (FA) at the same core position. The power in the IFA hottest plate is only 6.79 % higher than the correspondent IEA-R1 fuel plate. The IFA power distribution was also compared to the hottest FA of the core. The power of each IFA fuel plate was below its correspondent hottest FA fuel plate. In addition, the total IFA power is 18.40 % less than the hottest FA in the core. No significant power peaks occur in the IFA during operation. As future works, thermal–hydraulic calculations will be performed considering this calculated power distribution and no hot spots are expected.
Análise de temperaturas em um elemento combustível do reator de pesquisas IEA-R1 durante evento de perda lenta de vazão com RELAP
2023 - CAMPOS, ROGERIO C. de; BELCHIOR JUNIOR, ANTONIO; SOARES, HUMBERTO V.; UMBEHAUN, PEDRO E.; TORRES, WALMIR M.; ANDRADE, DELVONEI A. de
O código RELAP (Reactor Excursion and Leak Analysis Program) é amplamente utilizado para realizar análises de acidentes em reatores nucleares de potência ou de pesquisa. O presente trabalho apresenta uma simulação do transiente de perda lenta de vazão no núcleo do reator a partir de um modelo com RELAP para o reator de pesquisas IEA-R1 contemplando a piscina, o núcleo do reator, toda tubulação e válvulas do circuito primário, o tanque de decaimento, bomba de circulação principal, trocador de calor e tubulação de retorno à piscina. A modelagem proposta conseguiu representar toda a fenomenologia do acidente, ou seja, o comportamento das temperaturas desde o início da perda de vazão, desligamento do reator, seguida da abertura da válvula de circulação natural até a reversão da direção do escoamento no núcleo do reator. A comparação com resultados experimentais mostrou diferenças de temperaturas de 2,3°C para o fluido e de até 4°C para o revestimento.
Computational fluid dynamics analysis of an open-pool nuclear research reactor core for fluid flow optimization using a channel box
2023 - SCURO, N.L.; ANGELO, G.; ANGELO, E.; PIRO, M.H.A.; UMBEHAUN, P.E.; TORRES, W.M.; ANDRADE, D.A.
A channel box installation in the IEA-R1 research reactor core was numerically investigated to increase fluid flow in fuel assemblies (FAs) and side water channels (SWCs) between FAs by minimizing bypasses in specific regions of the reactor core, which is expected to reduce temperatures and oxidation effects in lateral fuel plates (LFPs). To achieve this objective, an isothermal three-dimensional computational fluid dynamics model was created using Ansys CFX to analyze fluid flow distribution in the Brazilian IEA-R1 research reactor core. All regions of the core and realistic boundary conditions were considered, and a detailed mesh convergence study is presented. Results comparing both scenarios are presented in the percentage of use of the primary circuit pump. It is indicated that 21.4% of fluid bypass to unnecessary regions can be avoided with the channel box installation, which leads to the total mass flow from the primary circuit for all FAs increasing from 68.9% (without a channel box) to 77.6% (with a channel box). For the SWCs, responsible for cooling LFPs, an increment from 9.7% to 22.4%, avoiding all nondesired cross three-dimensional effects, was observed, resulting in a more homogeneous fluid flow and vertical velocities. It was concluded that the installation of a channel box numerically indicates an expressive mass flow increase and homogeneous fluid flow distribution for flow dynamics in relevant regions. This gives greater confidence to believe that lower temperatures, and consequently oxidation effects in LFPs, can be expected with a channel box installation.
RANS-based CFD calculation for pressure drop and mass flow rate distribution in an MTR fuel assembly
2021 - SCURO, N.L.; ANGELO, G.; ANGELO, E.; UMBEHAUN, P.E.; TORRES, W.M.; SANTOS, P.H.G.; FREIRE, L.O.; ANDRADE, D.A.
This work presents a Reynolds-averaged Navier Stokes–based computational fluid dynamics methodology for the calculation of pressure drop and mass flow rate distribution in a material test reactor flat-plate-type standard fuel assembly (SFA) of the IEA-R1 Brazilian research reactor to predict future improvements in newer SFA designs. The results improve the understanding of the origin of fuel plate oxidation due to high temperatures, and consequently, due to the internal flow dynamics. All numerical analyses were performed with the ANSYS-CFX® commercial code. The observed results show that the movement pin decreases the central channel mass flow due to the length of the vortex at the inlet region. However, the outlet nozzle showed greater general influence in the flow dynamics. It should have a more gradual cross-section transition being away from the fuel plates or a squarer-shaped design to get a more homogeneous mass flow distribution. Optimizing both regions could lead to a better cooling condition. The validation of the IEA-R1 numerical methodology was made by comparing the McMaster University’s dummy model experiment with a numerical model that uses the same numerical methodology. The experimental data were obtained with laser Doppler velocimetry, and the comparison showed good agreement for both pressure drop and mass flow rate distribution using the Standard k-ω turbulence model.
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.
Two-phase flow void fraction estimation based on bubble image segmentation using Randomized Hough Transform with Neural Network (RHTN)
2020 - SERRA, PEDRO L.S.; MASOTTI, PAULO H.F.; ROCHA, MARCELO S.; ANDRADE, DELVONEI A. de; TORRES, WALMIR M.; MESQUITA, ROBERTO N. de
The International Atomic Energy Agency (IAEA) has been encouraging the use of passive cooling systems in new designs of nuclear power plants. Next nuclear reactor generations are intended to have simpler and robust safety resources. Natural Circulation based systems hold an undoubtedly prominent position among these. The study of limiting conditions of these systems has led to instability behavior analysis where many different two-phase flow patterns are present. Void fraction is a key parameter in thermal transfer analysis of these flow instability conditions. This work presents a new method to estimate void fraction from images captured of an experimental two-phase flow circuit. The method integrates a set of Artificial Neural Networks with a modified Randomized Hough Transform to make multiple scans over acquired images, using crescent-sized masks. This method was called Randomized Hough Transform with Neural Network (RHTN). Each different mask size is chosen according with bubble sizes, which are the main ‘objects of interest’ in this image analysis. Images are segmented using fuzzy inference with different parameters adjusted based on acquisition focus. Void fraction calculation considers the volume of the imaged geometrical section of flow inside cylindrical glass tubes considering the acquisition depth-of-field used. The bubble volume is estimated based on geometrical parameters inferred for each detected bubble. The image database is obtained from experiments performed on a vertical two-phase flow circuit made of cylindrical glass where flow-patterns visualization is possible. The results have shown that the estimation method had good agreement with increasing void fraction experimental values. RHTN has been very efficient as bubble detector with very low ‘false-positive’ cases (< 0.004%) due robustness obtained through integration between Artificial Neural Networks with Randomized Hough Transforms.
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.
Two-phase flow bubble detection method applied to natural circulation system using fuzzy image processing
2018 - BUENO, R.C.; MASOTTI, P.H.F.; JUSTO, J.F.; ANDRADE, D.A.; ROCHA, M.S.; TORRES, W.M.; MESQUITA, R.N. de
Natural 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.

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