NIKOLAS LYMBERIS SCURO

NIKOLAS LYMBERIS SCURO

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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.
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.
A CFD analysis of the flow dynamics of a directly-operated safety relief valve
2018 - SCURO, N.L.; ANGELO, E.; ANGELO, G.; ANDRADE, D.A.
A three-dimensional numerical study on steady state was designed for a safety relief valve using several openings and inlet pressures. The ANSYS-CFX (R) commercial code was used as a CFD tool to obtain several properties using dry saturated steam revised by IAPWS-IF97. Mass flow and discharge coefficient calculated from simulations are compared to the ASME 2011a Section 1 standard. The model presented constant behavior for opening lifts smaller than 12mm and is very reasonable when compared to the standard (ASME). In addition, the conventional procedure to design normal disc force assumes that all the fluid mechanical energy was converted into work; however, the CFD simulations showed that average normal disc force is about 19% lower than theoretical ASME force, which could prevent the valve oversizing. A numerical validation was conducted for a transonic air flow through a converging-diverging diffuser geometry to verify the solver's ability to capture the position and intensity of a shockwave: the results showed good agreement with the benchmark experiments.
CFD analysis of blockage length on a partially blocked fuel rod
2017 - SCURO, N.L.; ANGELO, G.; ANGELO, E.; ANDRADE, D.A.
In LOCA accidents, fuel rods may balloon by the increasing of pressure difference between fuel rod and core vessel. With the balloon effect, the swelling can partially block the flow channel, affecting the coolability during reflood phase. In order to analyze the influence of blockage length after LOCA events, many numerical simulations using Ansys-CFX code have been done in steady state condition, characterizing the final phase of reflood. Peaks of temperature are observed in the middle of the fuel rod, followed by a temperature drop. This effect is justified by the increasing of heat transfer coefficient, originated from the high turbulence effects. Therefore, this paper considers a radial blockage of 90%, varying just the blockage length. This study observed that, for the same boundary conditions, the longer the blockage length originated after LOCA events, the higher are the central temperatures in the fuel rod.
CFD analysis of blockage length on a partially blocked fuel rod
2017 - SCURO, NIKOLAS L.; ANGELO, GABRIEL; ANGELO, EDVALDO; ANDRADE, DELVONEI A. de
In LOCA accidents, fuel rods may balloon by the increasing of pressure difference between fuel rod and core vessel. With the balloon effect, the swelling can partially block the flow channel, affecting the coolability during reflood phase. In order to analyze the influence of blockage length after LOCA events, many numerical simulations using Ansys-CFX code have been done in steady state condition, characterizing the final phase of reflood. Peaks of temperature are observed in the middle of the fuel rod, followed by a temperature drop. This effect is justified by the increasing of heat transfer coefficient, originated from the high turbulence effects. Therefore, this paper considers a radial blockage of 90%, varying just the blockage length. This study observed that, for the same boundary conditions, the longer the blockage length originated after LOCA events, the higher are the central temperatures in the fuel rod.
Experimental investigation of critical velocity in a parallel plate research reactor fuel assembly
2017 - CASTRO, ALFREDO J.A.; SCURO, NIKOLAS L.; ANDRADE, DELVONEI A.
The fuel elements of a MTR (Material Testing Reactor) type nuclear reactor are mostly composed of aluminumcoated fuel plates containing the core of uranium silica (U3Si2) dispersed in an aluminum matrix. These plates have a thickness of the order of millimeters and are much longer in relation to their thickness. They are arranged in parallel in the assembly of the fuel element to form channels between them a few millimeters in thickness, through which there is a flow of the coolant. This configuration, combined with the need for a flow at high flow rates to ensure the cooling of the fuel element in operation, may create problems of mechanical failure of fuel plate due to the vibration induced by the flow in the channels. In the case of critical velocity excessive permanent deflections of the plates can cause blockage of the flow channel in the reactor core and lead to overheating in the plates. For this study an experimental bench capable of high volume flows and a test section that simulates a plate-like fuel element with three cooling channels were developed. The dimensions of the test section were based on the dimensions of the Fuel Element of the Brazilian Multipurpose Reactor (RMB), whose project is being coordinated by the National Commission of Nuclear Energy (CNEN). The experiments performed attained the objective of reaching Miller's critical velocity condition. The critical velocity was reached with 14.5 m/s leading to the consequent plastic deformation of the flow channel plates.