This study focuses on the investigation of flow behavior at low Reynolds numbers by the experimental and numerical performance testing of micro-channel heat exchangers. An experimental study of the heat transfers and pressure drop of compact heat exchangers with louvered fins and flat tubes was conducted within a low air-side Reynolds number range of 20 < ReLp < 225. Using an existing low-speed wind tunnel, 26 sample heat exchangers of corrugated louver fin type, were tested. New correlations for Colburn j and Fanning friction f factor have been developed in terms of non-dimensional parameters. Within the investigated parameter ranges, it seems that both the j and f factors are better represented by two correlations in two flow regimes (one for ReLp = 20 – 80 and one for ReLp = 80 – 200) than a single regime correlation in the power-law format. The results support the conclusion that airflow and heat transfer at very low Reynolds numbers behaves differently from that at higher Reynolds numbers. The effect of the geometrical parameters on the heat exchanger performance was investigated. The numerical investigation was conducted for further understanding of the flow behavior at the range of experimentally tested Reynolds number. Ten different heat exchanger geometries with varied geometrical parameters obtained for the experimental studies were considered for the numerical investigation. The variations in the louver angle were the basis of the selection. The heat transfer and pressure drop performance was numerically investigated and the effect of the geometrical parameters was evaluated. Numerical results were compared against the experimental results. From the comparison, it is found that the current numerical viscous laminar models do not reflect experimentally observed transitional two regime flow behavior from fin directed to the louver directed at very low Reynolds number ranging from 20 to 200. The flow distribution through the fin and the louver region was quantified in terms of flow efficiency. The flow regime change was observed at very low Reynolds number similar to the experimental observations. However, the effect of two regime flow change does not reflect on the thermal hydraulic performance of numerical models. New correlations for the flow efficiency � have developed in terms of non-dimensional parameters.

Orientador: Prof. Dr. Pedro Carajilescov / Dissertação (mestrado) - Universidade Federal do ABC. Programa de Pós-Graduação em Energia, Santo André, 2016. / A utilização de elementos combustíveis do tipo placa, em reatores nucleares, está associada principalmente a reatores de pesquisa e reatores de propulsão naval (navios porta aviões e submarinos), trazendo benefícios imediatos na segurança e no desempenho termo-hidráulico do reator. Códigos computacionais são utilizados para o cálculo do comportamento termo-hidráulico do núcleo. Este trabalho apresenta o desenvolvimento de um código termo-hidráulico para um reator nuclear com elementos combustíveis na forma de placas, em linguagem de programação FORTRAN. A partir da entrada dos parâmetros geométricos e das condições de operação e de contorno do reator, o programa realiza a análise do escoamento em regime permanente de potência ou vazão por meio da solução das equações de conservação de massa, quantidade de movimento e energia; além disso, calcula o mínimo DNBR baseado na análise do canal crítico (faz uma análise de fluxo crítico de calor). O código aumenta a representação da malha radial usando o método da cadeia, realizando os cálculos em duas etapas: na Etapa 1, o núcleo é subdividido em subcanais cujo tamanho é equivalente a um elemento combustível e na Etapa 2, o elemento combustível quente é subdividido em subcanais de tamanho equivalente aos canais que o compõem. Na validação do programa, considerou-se o reator de pesquisa CARR (China Advance Research Reactor) e o reator do LABGENE (Reator brasileiro de propulsão naval), obtendo informações detalhadas do núcleo do reator como a perda de carga, distribuição de fluxo mássico, variação de velocidade e temperatura do escoamento ao longo dos canais, título termodinâmico e fluxo crítico de calor no canal quente. A análise mostrou bons resultados quando verificado frente aos obtidos para o reator CARR e para um típico reator de potência PWR. / The use of plate-type fuel assembly, in nuclear reactors, are mostly associated to researched reactors and naval propulsion reactors (aircraft carriers and submarines), bringing immediate benefits in security and thermal-hydraulic performance of the reactor. Computational codes are used to calculating the thermal-hydraulic core behavior. This project shows the development of thermal-hydraulic code for plate type fuel reactor, written whit FORTRAN programming language. According to geometric input data, operational and boundary conditions, the code involves the analysis of permanent regime of flow and power through the solution of mass, momentum and energy conservation equation; Furthermore, it makes the calculation of minimum DNBR, based on an analysis of critical channel (Making an analysis of the maximum heat flux). The code has maximized the radial mesh with the use of the chain or cascade method for two stages: in the first stage, the core is subdivided in sub channels, with size equivalent to a fuel assembly and the stage two, the hot fuel assembly is subdivided in sub channels with size equivalent to the one channel that comprise. For the validation of program, was considered the research reactor CARR (China Advance Research Reactor), and the LABGENE reactor (Brazilian reactor of naval propulsion), getting detail information of reactor core as the change of the static pressure in the channel, flux distribution, variation of coolant temperature and coolant velocities, quality and local flux heat in the critical channel. The analysis showed good agreement when checked with the results obtained for CARR reactor and for a typical reactor power PWR.

The aim of the thesis is to design a heat exchanger for heating natural gas in the mining process and subsequent treatment. The first part describes natural gas itself, its origin, mining and ways of treatment. There are described technologies used for finding the deposits itself or processes by which water, sulfur and solid particles are removed from the extracted gas. The next section deals with the types of heat exchangers most commonly used in industry. There are also described basic relations for heat-hydraulic calculation of heat exchanger. In the practical part the heat exchanger itself was designed. World-wide HTRI software was used for the heat-hydraulic calculation and for strength calculation according to the standard ČSN EN 13445 was used sofware Sant´ Ambrogio.

This graduation thesis is concerned with a thermal exchange unit issue whereof one working substance complies with a two-phase mode of a flow. This unit is made for the steam generation. The thesis is supposed to solve operation problems causes of the given unit and to make a proposal of an appropriate solution that is supported by performed calculations. Part of the the work is strength calculation. This work also introduces the thermal-hydraulic processes issue of the steam generation equipment. There are also performed thermal-hydraulic control calculations in consideration of newly set-up operation parameters of the given equipment‘s working substances.

The proposed master´s thesis looks into the aspects of tubular heat exchangers modeling using available software tools. In the first – theoretical part there is a description of distribution and types of heat exchangers, including a detailed description of the industrial heat exchanger solved in this thesis. Another chapter is devoted to the main computational relations and calculation methods used in the context of thermo-hydraulic calculation. The main part of thesis consists of chapters, which are devoted to selected available software and perform calculations using the software in the specific industrial case. Finally, there is summary of the various software and a recommendation for the specific program for the need.

Após o aumento de potência do reator IEA-R1 de 2 MW para 5 MW observou-se um aumento da taxa de corrosão nas placas laterais de alguns elementos combustíveis e algumas dúvidas surgiram com relação ao valor de vazão utilizada nas análises termo-hidráulicas. A fim de esclarecer e medir a distribuição de vazão real pelos elementos combustíveis que compõe o núcleo do reator IEA-R1, um elemento combustível protótipo, sem material nuclear, chamado DMPV-01 (Dispositivo para Medida de Pressão e Vazão), em escala real, foi projetado e construído em alumínio. A vazão no canal entre dois elementos combustíveis é muito difícil de estimar ou ser medida. Esta vazão é muito importante no processo de resfriamento das placas laterais. Este trabalho apresenta a concepção e construção de um elemento combustível instrumentado para medir a temperatura real nestas placas laterais para melhor avaliar as condições de resfriamento do combustível. Quatorze termopares foram instalados neste elemento combustível instrumentado. Quatro termopares em cada canal lateral e quatro no canal central, além de um termopar no bocal de entrada e outro no bocal de saída do elemento. Existem três termopares para medida de temperatura do revestimento e um para a temperatura do fluido em cada canal. Três séries de experimentos, para três configurações distintas, foram realizadas com o elemento combustível instrumentado. Em dois experimentos uma caixa de alumínio foi instalada ao redor do núcleo para reduzir o escoamento transverso entre os elementos combustíveis e medir o impacto na temperatura das placas externas. Dada a tamanha quantidade de informações obtidas e sua utilidade no projeto, melhoria e capacitação na construção, montagem e fabricação de elementos combustíveis instrumentados, este projeto constitui um importante marco no estudo de núcleos de reatores de pesquisa. As soluções propostas podem ser amplamente utilizadas para outros reatores de pesquisa. / After the IEA-R1 upgrade from 2 MW to 5 MW it was observed that the corrosion rate increased in a lateral plate of one fuel element and some issues appeared concerning the flow values used in the thermal-hydraulic analysis. In order to clear it up and measure the actual flow distribution among the fuel elements composing the IEA-R1 active core, a dummy element without nuclear fuel material, called DMPV-01 (Pressure and Flow Measurement Device), full scale, was designed and manufactured in aluminum. The flow rate in the channel between two fuel assemblies is very difficult to estimate or measure. This flow rate is very important to the cooling process of the external plates. This work presents the design and construction of an instrumented fuel assembly in order to measure the actual temperature in these lateral plates. Fourteen thermocouples were installed in this instrumented fuel assembly. Four in each lateral channel, one in the inlet nozzle and one in the outlet nozzle. There are three thermocouples in each channel to measure the clad temperature and one thermocouple to measure the fluid temperature. Three series of experiments, for three different core configuration were carried out with the instrumented fuel assembly. In two experiments a box was installed around the core to reduce the cross flow between the fuel assembly and measure the impact in the temperatures of external plates. Given the amount of information generated and its utility in the design, improvement and qualification in construction, assembly and manufacturing of instrumented fuel, this project turned out to be an important landmark on the thermal-hydraulic study of research reactor cores. The proposed solutions could be useful for other research reactors.

Após o aumento de potência do reator IEA-R1 de 2 MW para 5 MW observou-se um aumento da taxa de corrosão nas placas laterais de alguns elementos combustíveis e algumas dúvidas surgiram com relação ao valor de vazão utilizada nas análises termo-hidráulicas. A fim de esclarecer e medir a distribuição de vazão real pelos elementos combustíveis que compõe o núcleo do reator IEA-R1, um elemento combustível protótipo, sem material nuclear, chamado DMPV-01 (Dispositivo para Medida de Pressão e Vazão), em escala real, foi projetado e construído em alumínio. A vazão no canal entre dois elementos combustíveis é muito difícil de estimar ou ser medida. Esta vazão é muito importante no processo de resfriamento das placas laterais. Este trabalho apresenta a concepção e construção de um elemento combustível instrumentado para medir a temperatura real nestas placas laterais para melhor avaliar as condições de resfriamento do combustível. Quatorze termopares foram instalados neste elemento combustível instrumentado. Quatro termopares em cada canal lateral e quatro no canal central, além de um termopar no bocal de entrada e outro no bocal de saída do elemento. Existem três termopares para medida de temperatura do revestimento e um para a temperatura do fluido em cada canal. Três séries de experimentos, para três configurações distintas, foram realizadas com o elemento combustível instrumentado. Em dois experimentos uma caixa de alumínio foi instalada ao redor do núcleo para reduzir o escoamento transverso entre os elementos combustíveis e medir o impacto na temperatura das placas externas. Dada a tamanha quantidade de informações obtidas e sua utilidade no projeto, melhoria e capacitação na construção, montagem e fabricação de elementos combustíveis instrumentados, este projeto constitui um importante marco no estudo de núcleos de reatores de pesquisa. As soluções propostas podem ser amplamente utilizadas para outros reatores de pesquisa. / After the IEA-R1 upgrade from 2 MW to 5 MW it was observed that the corrosion rate increased in a lateral plate of one fuel element and some issues appeared concerning the flow values used in the thermal-hydraulic analysis. In order to clear it up and measure the actual flow distribution among the fuel elements composing the IEA-R1 active core, a dummy element without nuclear fuel material, called DMPV-01 (Pressure and Flow Measurement Device), full scale, was designed and manufactured in aluminum. The flow rate in the channel between two fuel assemblies is very difficult to estimate or measure. This flow rate is very important to the cooling process of the external plates. This work presents the design and construction of an instrumented fuel assembly in order to measure the actual temperature in these lateral plates. Fourteen thermocouples were installed in this instrumented fuel assembly. Four in each lateral channel, one in the inlet nozzle and one in the outlet nozzle. There are three thermocouples in each channel to measure the clad temperature and one thermocouple to measure the fluid temperature. Three series of experiments, for three different core configuration were carried out with the instrumented fuel assembly. In two experiments a box was installed around the core to reduce the cross flow between the fuel assembly and measure the impact in the temperatures of external plates. Given the amount of information generated and its utility in the design, improvement and qualification in construction, assembly and manufacturing of instrumented fuel, this project turned out to be an important landmark on the thermal-hydraulic study of research reactor cores. The proposed solutions could be useful for other research reactors.

Les boucles diphasiques à pompage thermo-capillaire de type LHP (pour Loop Heat Pipe, en anglais), dont le fonctionnement s’apparente à celui d’un caloduc, permettent un transfert de chaleur particulièrement efficace et entièrement passif entre une source chaude et une source froide. Ce transfert s’effectue au moyen d’un fluide diphasique, mû grâce à la force motrice capillaire générée par un matériau poreux contenu dans l’évaporateur/réservoir de la LHP. Outre son rôle de barrière hydraulique entre les phases liquide et vapeur, ce matériau doit assurer une fonction de barrière thermique afin de favoriser l’évaporation du liquide. L’aptitude du matériau à remplir ses fonctions dépend étroitement de sa microstructure, elle-même liée à la méthode de fabrication. Dès lors, l’objectif de la thèse est d’associer la science des matériaux à celle de la thermique, pour améliorer les procédures de fabrication de structures capillaires existantes ou tester de nouvelles méthodes, et aboutir à des structures dont les caractéristiques sont en adéquation avec celles qui sont recherchées. / The capillary pumped loops (CPL) or loop heat pipes (LHP), whom the operating principle is similar to classic heat pipes, enable an efficient heat transfer between a hot source and a cold source without additional energy sources. Indeed, a porous structure provides a capillary force that enables a two-phase fluid to circulate around the loop, transferring the heat from the evaporator to the condenser. The porous structure acts as a hydraulic barrier between the two phases and as a thermal barrier enabling the liquid evaporation. The ability of the capillary structure to fulfil its mission depends on its microstructure, and thus on the manufacturing process. Therefore, the objective of the present thesis is to join the thermal sciences with the material sciences in order to improve the existing manufacturing procedure or even to test new ones. It aims at obtaining capillary structures corresponding to heat transfer applications.

Advanced computational tools are applied to simulate a nuclear power plant (NPP) control rod assembly ejection (CRE) accident. The impact of these reactivity-initiated accidents (RIAs) on core reactivity behavior, 3D power distribution and stochastic reactivity estimation are evaluated. The three tools used are: the thermal-hydraulic (TH) RELAP5 (R5) code, the neutronic (NK) PARCS (P3D) code, and the coupled version P3D/R5, with specially developed linkage using the environment code MATLAB. This study considers three different-size cores: NPP1 (2772 MWt); NPP2 (530 MWt); and NPP3 (1061 MWt). The three cores have the same general design and control rod assembly (CRA) positions, and the ejected CRA has similar worth and at the same rod ejection pace. The CRE is assessed under both hot zero power (HZP) and hot full power (HFP) conditions. The analyses indicate that RIA modeling and simulation should be carried out through a systematic coding and configuration approaches, otherwise the results will not capture the true transient behavior of the core under analysis. The simulation of one code depends on the appropriate configuration of parameters generated by the other code and on the correct determination of the TH/NK mapping weight factors for the various mesh regions in each of the models. From the design point of view, the standalone codes predict milder magnitude of power and reactivity increase compared to the coupled P3D/R5 simulation. The magnitudes of reduced peak power and reactivity become larger as the core size shrinks. The HFP simulation shows that the three NPPs have the same transient peak value, but the post-transient steady power is lower for a smaller core. The HZP analysis indicates that the transient peak is lower for the smaller core, but the post-transient power occurs at the same level. The three-dimensional (3D) power distributions are different among the HFP and HZP cases, but do not depend on the size of the core. The results indicate: i) HFP: core power increases in the area surrounding the ejected rod/bank assembly, and this increase becomes lower as the NPPs shrinks however, the power is well-distributed after the transient; and ii) HZP: the area surrounding the CRA stays hotter, but the 3D peak assembly factor becomes lower, during and after the transients, as the NPPs shrinks. These features confirm that the smaller cores yield a safer response to a given inserted reactivity compared to larger cores. A stochastic extended Kalman filter (EKF) algorithm is implemented to estimate the reactivity based on the reactor power profile, after the addition of random noise. The inverse point kinetics (IPK) deterministic method is also implemented and the results of the application of EKF and IPK are compared to the P3D/R5 simulation. The following sophisticated strategies made the EKF algorithm robust and accurate: the system is modeled by a set of continuous time nonlinear stochastic differential equations; the code uses a time step directly based on the power measured and applies that to the model for online discretization and linearization; filter tuning goes automatically up from the first time step; and the state noise covariance matrix is updated online at each time step. It was found that the IPK reactivity has higher noise content compared to the EKF reactivity for all cases. Thus, the EKF presents superior and more accurate results. Furthermore, under a small reactivity insertion, the IPK reactivity varies widely from positive to negative values: this variation is not observed within the EKF. A sensitivity analysis for three distinct standard deviation (SD) noise measurements suggests that EKF is superior to IPK method, independent of the noise load magnitude. As the noise content increases, the error between the IPK and P3D/R5 reactivity also increases. A sensitivity analysis for five distinct carry-over effects of different random noise loads indicates that the random addition of different noise loads to the reactor power does not change the overall performance of both algorithms. / Este trabalho aplica métodos computacionais avançados para simular a ejeção de barras de controle (CRE) em uma planta térmica nuclear (NPP). São avaliados o impacto da ocorrência de acidentes iniciados por reatividade (RIAs) na reatividade total, na distribuição da potência em três dimensões (3D) e na determinação da reatividade. As ferramentas utilizadas são: o código termo-hidráulico (TH) RELAP5 (R5), o código neutrônico (NK) PARCS (P3D), a versão acoplada P3D/R5, e o ambiente computacional MATLAB. Este estudo considera três reatores nucleares de diferentes tamanhos: NPP1 (2772 MWT); NPP2 (530 MWt); e NPP3 (1061 MWt). Os três núcleos possuem projeto similar e idêntica posição dos grupos das barras de controle (CRA), além do mesmo valor de reatividade diferencial das CRA ejetadas e idêntica velocidade de ejeção. A ocorrência da CRE é avaliada sob condições de hot zero power (HZP) e de hot full power (HFP). As análises indicam que a modelagem e a simulação de RIAs devem ser realizadas sistematicamente, caso contrário os resultados não irão refletir o comportamento em regime transitório do núcleo. A simulação de um modelo em um código depende da apropriada configuração de parâmetros gerados pelo outro código e da determinação adequada do mapeamento TH/NK para as várias malhas dos modelos. Do ponto de vista de projeto, a utilização de códigos independentes resulta em cálculos de potência e reatividade conservadores em comparação com os resultados utilizando-se P3D/R5. Os picos de potência e de reatividade são menores à medida que o núcleo encolhe. A simulação em condições de HFP resulta em valores de pico de potência similares durante transitório para as três NPPs, mas a potência de pós-transitórios é menor para o menor núcleo. A análise em condições de HZP também indica que o valor máximo durante o transitório é menor para o menor núcleo, mas o pós-transitórios ocorre aos mesmos níveis de potência das demais NPPS. A distribuição de potência em 3D também apresenta resultados distintos para condições de HFP e HZP, mas tais resultados são independentes do tamanho do núcleo: i) HFP: há um aumento da potência do núcleo em torno da CRE, mas tal comportamento diminui para núcleos menores - no entanto, a potência é bem distribuída após o transitório; e ii) HZP: há aumento de potência na área do CRE, mas o pico de potência em 3D é menor durante e depois dos transitórios para núcleos menores. Tais características indicam que os núcleos menores respondem de forma mais segura quando da inserção de reatividade em comparação a reatores de maiores dimensões. O método estocástico de filtragem de Kalman estendido (EKF) foi codificado para estimar a reatividade com base no perfil de potência da NPP, após a adição de ruído aleatório. O método determinístico da cinética pontual inversa (IPK) também foi implementado e os resultados da aplicação dos algoritmos do EKF e IPK foram comparados com os resultados da simulação do P3D/R5. As seguintes estratégias, implementadas neste trabalho, possibilitaram a aplicação robusta e precisa do EKF: o sistema foi modelado por um conjunto de equações diferenciais não-lineares estocásticas de tempo contínuo; o algoritmo obtém o passo de tempo diretamente da potência medida e aplica-o ao modelo para a discretização e linearização online; o ajuste do filtro ocorre automaticamente a partir do primeiro passo de tempo; e a matriz de covariância do ruído no estado é atualizada online. Verificou-se que a reatividade calculada pelo método IPK possui maior nível de ruído quando comparada ao EKF para todos os casos estudados. Portanto, o EKF apresenta resultados superiores e mais precisos. Além disso, sob uma pequena inserção de reatividade, a reatividade calculada pelo método IPK varia consideravelmente de valores positivos para negativos: esta variação não é observada com o EKF. Uma análise de sensibilidade para três desvios padrão (SD) sugere que o algoritmo EKF é superior ao método IPK, independente da magnitude do ruído. Com o aumento da magnitude do ruído, o erro entre as reatividades calculadas pelo IPK e pelo P3D/R5 aumenta. A análise de sensibilidade para cinco ruídos aleatórios indica que a adição de ruído na potência do reator não altera o desempenho global de ambos os algoritmos.

The Multi-Application Small Light Water Reactor (MASLWR) is a small natural circulation pressurized light water reactor design that was developed by Oregon State University (OSU) and Idaho National Engineering and Environmental Laboratory (INEEL) under the Nuclear Energy Research Initiative (NERI) program to address the growing demand for energy and electricity. The MASLWR design is geared toward providing electricity to small communities in remote locations in developing countries where constructions of large nuclear power plants are not economical. The MASLWR reactor is designed to operate for five years without refueling and with fuel enrichment up to 8 %. In 2003, an experimental thermal hydraulic research facility also known as the OSU MASLWR Test Facility was constructed at Oregon State University to examined the performance of new reactor design and natural circulation reactor design concepts. This thesis is focused on the thermal hydraulics analysis and fuel performance analysis of the MASLWR prototypical cores with fuel enrichment of 4.25 % and 8 %. The goals of the thermal hydraulic analyses were to calculate the departure nucleate boiling ratio (DNBR) values, coolant temperature, cladding temperature and fuel temperature profiles in the hot channel of the reactor cores. The thermal hydraulic analysis was performed for steady state operation of the MASLWR prototypical cores. VIPRE Version 01 is the code used for all the computational modeling of the prototypical cores during thermal hydraulic analysis. The hot channel and hot rod results are compared with thermal design limits to determine the feasibility of the prototypical cores. The second level of analysis was performed with a fuel performance code FRAPCON for the limiting MASLWR fuel rods identified by the neutronic and thermal hydraulic analyses. The goals of the fuel performance analyses were to calculate the oxide thickness on the cladding and fission gas release (FGR). The oxide thickness results are compared with the acceptable design limits for standard fuel rods. The results in this research can be helpful for future core designs of small light water reactors with natural circulation. / Graduation date: 2012