NIKOLAS LYMBERIS SCURO

NIKOLAS LYMBERIS SCURO

Página do item completo

Resultados de Busca

Agora exibindo 1 - 10 de 14

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.
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.
A simulation model for capacity planning of nuclear fuel plants for research reactors
2021 - NEGRO, M.L.N.; DURAZZO, M.; MESQUITA, M.A.; SCURO, N.L.; CARVALHO, E.F.U.; ANDRADE, D.A.
The demand for nuclear fuel for research reactors is increasing worldwide. However, some nuclear fuel factories have low production volumes. Literature regarding how to expand the capacity of those facilities in a safe and reliable way is scarce. Thus, the purpose of this work is to propose and validate a conceptual model for increasing the production capacity of such factories. The facilities addressed here are those that produce plate-type fuel elements loaded with LEU U3Si2-Al, which are typically used in nuclear research reactors. Data from a real nuclear fuel plant was collected and applied to the model, thus setting up a case study. Two different strategies, as well as several production scenarios, were conceived for the use of the model. Each scenario experiments with the different possibilities of enlarging capacity. Discrete events simulation was used in order to cover all production scenarios. The tests indicated significant increases in productive capacity, thus showing that the model fully achieved its proposed objective. One of the main conclusions to be highlighted is the model’s effectiveness, which was demonstrated by using the model in two different strategies and obtaining increases in capacity with both of them.
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.
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.
Simulação numérica de um acidente tipo perda lenta de vazão em um reator nuclear de pesquisa
2019 - SCURO, NIKOLAS L.
As simulações numéricas de acidentes em reatores nucleares de pesquisa necessitam de constante aprimoramento, originando metodologias validadas, o que permite aproximar os cálculos numéricos a um comportamento físico. O trabalho proposto consiste em elaborar uma metodologia numérica tridimensional para análise de um acidente tipo perda lenta de vazão, comumente nomeado de SLOFA, do inglês, slow loss of flow accident, para o reator nuclear IEA-R1. Utilizando códigos numéricos para escoamentos tridimensionais (ANSYS CFX®) foi possível observar a dinâmica do escoamento, prever a localização da temperatura máxima do revestimento e o instante da inversão do sentido de escoamento. Sete modelos de turbulência foram analisados individualmente para elaboração da metodologia, porém, inúmeras dificuldades foram observadas no processo de solução para os modelos ZE, EVTE, SSG, k - ε, k - ω, SST e DES. O modelo que atendeu aos requisitos estabelecidos, entre eles, tempo computacional e solução numérica compatível com solução física, foi o modelo de turbulência k - ω. Entre as justificativas para este resultado pode-se citar a ausência da lei logarítmica de parede e simplicidade na solução das equações de transporte para condição analisada. Os resultados apresentaram alinhamento quantitativo e qualitativo com as curvas de temperatura experimentais. Nas condições de regime permanente quanto para o regime transiente, o desvio máximo observado foi de 3,4°C para temperatura. As curvas de temperatura numérica capturam o mesmo comportamento físico observado nos testes experimentais, tanto no instante da inversão do escoamento, quanto no início da perda dos efeitos do empuxo. Portanto, esta metodologia tridimensional representa um avanço frente aos resultados apresentados pelos códigos unidimensionais reportados na literatura (RELAP, MERSAT, CATHARE) para a mesma base de dados experimental, visto que o desvio médio observado nestes códigos é de 7,2°C.
New formulation for semi-empirical correlations for penetration jets
2019 - PACHECO, R.R.; FREIRE, L.O.; ROCHA, M.S.; SCURO, N.L.; MENEZES, M.O.; ANDRADE, D.A.
Correlations for the extension of a water vapor jet injected in a liquid pool were historically proposed considering the mass flux (kg/m2/s) as a constant. The results were satisfactory, however adjusting the values by linear regression. Although, it presents the following drawbacks: 1) the formulation is only valid for the specific range of data for what it was created; 2) it does not allow the analytical evaluation of the heat transfer coefficient from the extension equation. This paper proposes a new formulation for the calculation of the mass flux, in such a way to remove both of these drawbacks.
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.