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This sensitivity analysis of the momentum and turbulence equations uses the Eulerian two-phase approach and intends to achieve a validated modeling for subcooled boiling flows. The studied regime is relevant to many industry applications and to the critical heat flux phenomenon, an important process for design and safety analyses of water-cooled nuclear power plants. Ascending water flows in heated circular pipes are modeled under subcooled boiling condition and applying a combination of computational multi-fluid dynamics (CMFD) models, each one for the estimation of specific terms of the governing equations. Turbulence closure, drag, lift, turbulent interaction, turbulent dispersion, wall lubrication and interfacial area models are investigated through variations around a reference model set defined based on the literature review, changing each model one at a time to study their impacts on the simulation results. Key parameters, such as void fraction and temperature profiles, are chosen for the evaluation of the results. Despite the difficulty of tracking the contributions of simultaneously applied models, the Legendre–Magnaudet lift force and the Burns et al. turbulent dispersion force models show good agreement with the experimental data of a flow at 4.5 MPa used for the quality verification of the simulations. These models are also applied for flows with pressures ranging from 1.5 to 15 MPa, and they lead to promising results. This study gives interesting insights on the influence of the distinct models composing momentum and turbulence calculations, contributing to the validation effort of the CMFD approach for subcooled boiling flows.

The heat transfer mechanism of natural convection has been extensively studied as a passive heat removal system of new nuclear power plants. Considering this aspect, the main objective of this study is to present an assessment of RELAP5 linear-equation solver under a transient two-fluid model for a two-phase natural circulation loop (NCL). For this assessment, three different approaches of linear-equation solvers for the hydrodynamic model are presented: the sparse matrix solver based on the Lower-Upper (LU) decomposition, the Border-Profile Lower Upper (BPLU) solver and the iterative method named Generalized Minimal Residual Method (GMRES). For comparison purposes, an experimental natural circulation loop made of glass tubes and using water as working fluid is analyzed. The onset of nucleate boiling observed during the experiment was predicted by all RELAP5 solvers as well as the representation of flow oscillations along the loop. Furthermore, it was noticed that the choice of the solver algorithm has a strong influence on the prediction of the two-phase natural circulation phenomena, since different wavelengths and amplitudes of flow instabilities were obtained for each approach.

Heat transfer processes using two-phase boiling flows are often found in the industry, due to the high-efficiency heat removal, with a minimum difference of temperature between the heated surface and fluid. Moreover, the use of computational fluid dynamics to solve safety issues related to nuclear reactors has increased rapidly, but a complete validation is still being carried out. Therefore, this study aims the assessment of different sub-models of the interfacial heat transfer coefficient which is used as a closure relation in the two-fluid multiphase model. As a manner to validate the numerical results, the set of experimental conditions of Bartolomei and Chanturiya (Therm Eng 14:123–128, 1967) were applied to an upwards flow in a circular channel, under high water saturation pressures. The interfacial sub-models were implemented into an axisymmetric simulation domain, using the Eulerian two-fluid approach. Three different saturation pressures and two uniform heat fluxes were considered in the simulation runs. Fixed mass flux and subcooled degree of 900 kg/m2 and 60 K were applied, respectively. A satisfactory agreement was achieved for the estimation of the heated wall temperature, the liquid bulk temperature and the onset of saturated boiling. Different heat transfer closure relations promoted different vapor volume fraction along the channel, indicating the importance of an adequate interfacial heat transfer correlation to predict the flow boiling phenomena.

This work contributes to enable and develop technologies to mount fast microreactors, to generate heat and electric energy, for the purpose to warm and to supply electrically spacecraft equipment and, also, the production of nuclear space propulsion effect. So, for this purpose, the Brayton Cycle demonstrates to be an optimum approach for space nuclear power. The Brayton thermal cycle gas has as characteristic to be a closed cycle, with two adiabatic processes and two isobaric processes. The components performing the cycle's processes are compressor, turbine, heat source, cold source and recuperator. Therefore, the working fluid’s mass flow runs the thermal cycle that converts thermal energy into electrical energy, able to use in spaces and land devices. The objective is numerically to model the Brayton thermal cycle gas on nominal operation with one turbomachine composed for a radial-inflow compressor and turbine of a 40.8 kWe Brayton Rotating Unit (BRU). The Brayton cycle numerical modeling is being performed with the program RELAP5-3D, version 4.3.4. The nominal operation uses as working fluid a mixture 40 g/mole He-Xe with a flow rate of 1.85 kg/s, shaft rotational speed of 45 krpm, compressor and turbine inlet temperature of 400 K and 1149 K, respectively, and compressor exit pressure 0.931 MPa. Then, the aim is to get physical corresponding data to operate each cycle component and the general cycle on this nominal operation.

The present study concerns a detailed analysis of flow boiling phenomena under high pressure systems using a two-fluid Eulerian approach provided by a Computational Fluid Dynamics (CFD) solver. For this purpose, a vertical heated pipe made of stainless steel with an internal diameter of 15.4 mm was considered as the modeled domain. Two different uniform heat fluxes and three saturation pressures were applied to the channel wall, whereas water mass flux of 900 kg/m2 s was considered for all simulation cases. The model was validated against a set of experimental data and results have indicated a promising use of the CFD technique for estimation of the wall temperature, the liquid bulk temperature and the location of the departure of nucleate boiling. Changes in factors applied in the modeling of the interfacial heat transfer coefficient and bubble departure frequency were suggested, allowing a better prediction of the void fraction along the heated channel. The commercial CFD solver FLUENT 14.5 was used for the model implementation.

When a certain void fraction value is reached in the two-phase flow regime, a problem occurs in the COBRA-EN code. This problem was observed in the drift-flux model option and interrupts code execution. Two solutions are proposed to solve the problem.

A segurança de instalações nucleares é uma preocupação mundial que tem crescido, sobretudo, após o acidente nuclear de Fukushima. O estudo de acidentes em reatores nucleares de pesquisa tal como o Acidente de Perda de Refrigerante (APR), considerado por muitas vezes um acidente base de projeto, é importante para garantir a integridade da instalação. O APR pode levar ao descobrimento parcial ou total do núcleo do reator e, como condição de segurança, deve-se garantir que haja a remoção do calor de decaimento dos elementos combustíveis. Esse trabalho teve o objetivo de realizar experimentos de descobrimento parcial e total no Elemento Combustível Instrumentado (ECI), construído no Instituto de Pesquisas Energética e Nucleares (IPEN), a fim de estudar os possíveis APRs em reatores de pesquisa. Uma seção de testes, denominada STAR, foi projetada e construída para simular os APRs. O ECI foi irradiado no núcleo do reator IEA-R1 (IPEN) e inserido na STAR, que ficou totalmente imersa na piscina do reator. No ECI, foram instalados termopares para medição das temperaturas do revestimento e do fluido em várias posições axiais e radiais. Foram realizados experimentos para cinco níveis de descobrimento do ECI, um total e quatro parciais, em duas condições distintas de calor de decaimento. Na análise dos resultados, verificou-se que os casos de descobrimento total foram os mais críticos, ou seja, as temperaturas do revestimento foram as maiores quando comparadas com os casos de descobrimentos parciais. Adicionalmente, foi realizada a simulação numérica de dois experimentos com o código RELAP5, cujos resultados demonstraram ótima concordância com os dos níveis experimentais, e temperaturas maiores que as experimentais. As máximas temperaturas do revestimento alcançadas em todos os experimentos ficaram bem abaixo da temperatura de empolamento do combustível, que é de 500°C. Assim, a STAR provou ser um aparato experimental seguro e confiável para a realização de experimentos de perda de refrigerante. The safety of nuclear facilities has been a growing global concern mainly after the Fukushima nuclear accident. Studies on nuclear research reactor accidents such as the Loss of Coolant Accident (LOCA), considered many times a design basis accident, are important for guaranteeing the integrity of the plant. A LOCA may lead to the partial or complete uncovering of the fuel assemblies and assured decay heat removal is a safety condition. This work aimed to perform partial and complete uncovering experiments in the Instrumented Fuel Assembly (IFA) designed at the Instituto de Pesquisas Energéticas e Nucleares (IPEN) in order to study possible LOCAs in research reactors. A test section for experimental simulation of Loss of Coolant Accident named STAR was designed and built. The IFA was irradiated in the IEA-R1 and installed in the STAR, which was totally immersed in the reactor pool. Thermocouples were installed in the IFA to measure the clad and fluid temperatures in several axial and radial positions. The experiments were carried out for five levels of uncovering of IFA, being one complete uncovering and four partial uncovering, in two different conditions of decay heat. In the results analysis was observed that the cases of complete uncovering of the IFA were the most critical ones, that is, those cases presented higher clad temperatures when compared with partial uncovering cases. Additionally, a numerical simulation of two experiments was carried out by using the RELAP 5 code. The numerical results showed an optimum agreement with the experimental levels results and greater than the experimental temperatures. The maximum clad temperatures reached in all experiments were quite below the fuel blister temperature, which is 500 °C. Therefore, the STAR has proven to be a safe and reliable experimental apparatus for conducting loss of coolant experiments.