When a pressurized water reactor is operated for a sufficiently long period of time, a small number of fuel rods will develop ruptures in their claddings. These defects will leak volatile fission products into the primary coolant, including radioactive iodine. During steady-state operation of the reactor a low level iodine activity is thus present in the coolant. Initiation of a down-power or up-power transient will result in a rapid climb in the activity of the iodine which peaks at a level much higher than the initial activity. After this time the activity levels out and then slowly begins to decay back to a new steady-state level. This phenomenon is termed "iodine spiking.” A physical model of this process is sought for explanatory and predictive purposes. A FORTRAN code is developed that solves a system of differential equations which describe the production and removal of iodine in the fuel, gap region, and primary coolant. As much physics as possible is employed but some complicated diffusion processes have led to the utilization of certain parametric results obtained from empirical data. Actual PWR spiking data is also employed for comparison and adjustment of the model. It is the goal of this project to be able to utilize the model for predictive analysis· during actual PWR operation so that a better understanding of iodine spiking behavior can be obtained. / Master of Science

Increased conversion ratios and burnup can be achieved by mechanically changing the fuel-to-water volume ratio of a reactor over the core lifetime. As the fuel-to-water ratio decreases, the neutron spectrum softens, thereby increasing core reactivity. Proposed mechanical spectral shift reactors utilize this concept. RFD-1, a 1-dimensional, 4-group code was developed to compute fuel burnup cycles for spectral shift reactors. The code calculates burnup for a triangular core lattice having a beginning fuel to water ratio as high as 1.30. Core shutdown occurs at a fuel to water ratio of 0.50. The microscopic cross sections were obtained through use of the VIM code and tabulated for use in RFD-1 as a function of fuel to water ratio and burnup time. The fission product group cross sections were developed using the VIM and TOAFEW codes. The flexibility of RFD-1 allows the user to study a wide variety of possible core configurations. Results of RFD-1 show that increased conversion and burnup, using lower initial enrichments than that of standard Pressurized Water Reactors, result for mechanical spectral shift designs. The next step is to study specific spectral shift designs in greater detail. The RFD-1 code could be improved primarily through refinements in its cross section data tables. / Master of Science

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

Thesis (Nucl.E.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1981. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / by James North Loomis. / Nucl.E.

Um dos parâmetros termo-hidráulicos de segurança mais importantes no projeto e operação de reatores a água pressurizada é o fluxo de calor crítico (FCC). O FCC ocorre quando se atinge uma região de instabilidade na mudança de mecanismo de transferência de calor de uma parede aquecida para um fluido, aumentado drasticamente a temperatura da parede. Transientes em um reator nuclear podem afetar a taxa de geração de calor ou a fluxo de refrigerante no núcleo, prejudicando a retirada de calor das varetas combustíveis. Conhecer o FCC nestas condições é essencial para evitar danos às varetas e, consequentemente, a liberação de material radioativo. O objetivo deste trabalho é analisar o FCC para o LABGENE (Laboratório de Geração Nucleoelétrica) por meio do planejamento experimental e da simulação de seções de teste em condições de operação utilizando o código COBRAIIIc/MIT-1 e a correlação EPRI para o FCC. Considerou-se primeiramente seções de teste 3×3 de dois tamanhos distintos e os resultados para 100 pontos experimentais foram mostrados por meio de superfícies de resposta, a fim de melhor visualizar e analisar o comportamento de FCC para cada condição. Dois pontos importantes são os valores máximo e mínimo do FCC encontrados. O valor máximo (1,038 MBtu/hr.ft2 ou 3,27 MW/m2) indica o fluxo de calor necessário para a realização dos experimentos e o mínimo (0,162 MBtu/hr.ft2 ou 0,51 MW/m2) indica a pior condição de operação, a qual estaria mais próxima do ponto de ebulição. As simulações e modificações no código foram verificadas utilizando o banco de dados da Universidade de Columbia. Foram selecionados 2718 pontos experimentais referentes a seções de teste 5×5 com perfil de potência uniforme. Os resultados foram apresentados pela razão entre o valor predito e o valor experimental (DNBR) e os limites de tolerância unilateral 95/95 foram calculados, estando dentro dos valores esperados. / One of the most important thermal-hydraulic safety parameters for pressurized water reactor design and operation is the critical heat flux (CHF). The CHF occurs when a region of instability reached in the change of heat transfer mechanism from a hot wall to a fluid is reached, dramatically increasing the wall temperature. Transients in a nuclear reactor can affect the heat generation rate or the coolant flow in the core, impairing the removal of heat from the fuel rods. Knowledge of the CHF on these conditions is essential to prevent fuel rod damages and therefore the release of radioactive material. The main goal of this work is to analyze the CHF for LABGENE (Nuclear-electrical Generation Laboratory) by an experimental design and test sections simulation in operating conditions by using COBRAIIIc/MIT-1 code and the EPRI correlation for CHF. 3x3 test sections were initially considered for two different heights and outcomes for 100 experimental points were shown by means of response surfaces in order to better visualize and analyze the behavior of CHF for each condition. Two important points are the maximum and minimum values of the CHF found. The maximum value (1.038 MW/m2 or 3.27 MBtu/hr.ft2) indicates the power required for the experiments and the minimum one (0.162 MBtu/hr.ft2 or 0.51 MW/m2) indicates the worst operation condition, which would be closer to the boiling point. Code simulations and modifications were verified using the CHF database of Columbia University. 2718 data points pertaining to test sections 5×5 with uniform power profile were selected. The results were presented by the ratio between the predicted value and the experimental value (DNBR) and the limits of unilateral tolerance 95/95 were calculated, being within the expected values.

Um dos parâmetros termo-hidráulicos de segurança mais importantes no projeto e operação de reatores a água pressurizada é o fluxo de calor crítico (FCC). O FCC ocorre quando se atinge uma região de instabilidade na mudança de mecanismo de transferência de calor de uma parede aquecida para um fluido, aumentado drasticamente a temperatura da parede. Transientes em um reator nuclear podem afetar a taxa de geração de calor ou a fluxo de refrigerante no núcleo, prejudicando a retirada de calor das varetas combustíveis. Conhecer o FCC nestas condições é essencial para evitar danos às varetas e, consequentemente, a liberação de material radioativo. O objetivo deste trabalho é analisar o FCC para o LABGENE (Laboratório de Geração Nucleoelétrica) por meio do planejamento experimental e da simulação de seções de teste em condições de operação utilizando o código COBRAIIIc/MIT-1 e a correlação EPRI para o FCC. Considerou-se primeiramente seções de teste 3×3 de dois tamanhos distintos e os resultados para 100 pontos experimentais foram mostrados por meio de superfícies de resposta, a fim de melhor visualizar e analisar o comportamento de FCC para cada condição. Dois pontos importantes são os valores máximo e mínimo do FCC encontrados. O valor máximo (1,038 MBtu/hr.ft2 ou 3,27 MW/m2) indica o fluxo de calor necessário para a realização dos experimentos e o mínimo (0,162 MBtu/hr.ft2 ou 0,51 MW/m2) indica a pior condição de operação, a qual estaria mais próxima do ponto de ebulição. As simulações e modificações no código foram verificadas utilizando o banco de dados da Universidade de Columbia. Foram selecionados 2718 pontos experimentais referentes a seções de teste 5×5 com perfil de potência uniforme. Os resultados foram apresentados pela razão entre o valor predito e o valor experimental (DNBR) e os limites de tolerância unilateral 95/95 foram calculados, estando dentro dos valores esperados. / One of the most important thermal-hydraulic safety parameters for pressurized water reactor design and operation is the critical heat flux (CHF). The CHF occurs when a region of instability reached in the change of heat transfer mechanism from a hot wall to a fluid is reached, dramatically increasing the wall temperature. Transients in a nuclear reactor can affect the heat generation rate or the coolant flow in the core, impairing the removal of heat from the fuel rods. Knowledge of the CHF on these conditions is essential to prevent fuel rod damages and therefore the release of radioactive material. The main goal of this work is to analyze the CHF for LABGENE (Nuclear-electrical Generation Laboratory) by an experimental design and test sections simulation in operating conditions by using COBRAIIIc/MIT-1 code and the EPRI correlation for CHF. 3x3 test sections were initially considered for two different heights and outcomes for 100 experimental points were shown by means of response surfaces in order to better visualize and analyze the behavior of CHF for each condition. Two important points are the maximum and minimum values of the CHF found. The maximum value (1.038 MW/m2 or 3.27 MBtu/hr.ft2) indicates the power required for the experiments and the minimum one (0.162 MBtu/hr.ft2 or 0.51 MW/m2) indicates the worst operation condition, which would be closer to the boiling point. Code simulations and modifications were verified using the CHF database of Columbia University. 2718 data points pertaining to test sections 5×5 with uniform power profile were selected. The results were presented by the ratio between the predicted value and the experimental value (DNBR) and the limits of unilateral tolerance 95/95 were calculated, being within the expected values.

Les aciers de construction au carbone manganèse (C-Mn) sont largement utilisés pour diverses applications mécaniques, et en particulier pour les tuyauteries de circuit secondaire des centrales nucléaires de type Réacteurs à Eau sous Pression (REP). La robustesse des composants des circuits sous pression des REP vis-à-vis de la fissuration doit être démontrée, tant au niveau de la conception que de l’exploitation. Les aciers au C-Mn sont sensibles au vieillissement sous déformation qui entraîne une chute importante de ductilité et de ténacité entre 150 et 350°C,températures de service des tuyauteries du circuit secondaire. Ce phénomène est dû à une interaction entre les atomes de solutés et les dislocations, et peut se traduire entre autres par une sensibilité négative de la contrainte à la vitesse de déformation, et des localisations de la déformation plastique (Lüders, Portevin – Le Chatelier). L’origine physique du vieillissement sous déformation a été beaucoup étudiée, surtout dans les métaux purs, en relation avec le phénomène Portevin-Le Chatelier (PLC), mais son influence sur les propriétés mécaniques et notamment la rupture reste très controversée. L’objectif de la thèse est de modéliser le comportement et la rupture d’un acier au C-Mn dans un large domaine de température compris entre 20 et 350°C, en tenant compte des phénomènes de vieillissement sous déformation, et en particulier des localisations de déformation. Le comportement et la rupture de l’acier au C-Mn étudié ont été caractérisés expérimentalement dans le domaine 20-350°C à l’aide d’essais de traction sur éprouvettes lisses, sur éprouvettes axisymétriques entaillées, et d’essais de déchirure sur éprouvettes CT. Le modèle d’Estrin Kubin McCormick, prenant en compte le vieillissement sous déformation, a été identifié dans cette même gamme de température et la plupart des effets du vieillissement sous déformation ont pu être simulés numériquement : sensibilité négative de la contrainte d’écoulement à la vitesse de déformation, bandes de Lüders, effet PLC, modification des propriétés mécaniques de traction… Le modèle ainsi identifié a été appliqué à l’étude de la rupture d’éprouvettes lisses, entaillées et CT. La baisse de l’allongement réparti est bien décrite en traction sur éprouvettes lisses. Pour prévoir la rupture des éprouvettes entaillées, l’approche locale de la rupture a été appliquée (modèle de Rice et Tracey). Cette étude a donc permis de disposer d’un modèle prenant en compte le vieillissement sous déformation de 20°C à 350°C et décrivant les localisations de déformation plastique de type Lüdersou PLC, pour différentes géométries d’éprouvettes. Ce modèle a été utilisé pour simuler la rupture des aciers au C-Mn, suscitant ainsi une vision nouvelle pour comprendre la baisse de ductilité associée au vieillissement dynamique. / Pas de résumé en anglais disponible.

Orientadores: Elizabete Jordão, Moysés Alberto Navarro / Tese (doutorado) - Universidade Estadual de Campinas, Faculdade de Engenharia Química / Made available in DSpace on 2018-08-20T11:58:32Z (GMT). No. of bitstreams: 1 Rezende_HugoCesar_D.pdf: 10614958 bytes, checksum: aa2b2ce9d9ec481bd44e74cc217832a2 (MD5) Previous issue date: 2012 / Resumo: O escoamento monofásico termicamente estratificado ocorre em tubulações horizontais onde duas camadas diferentes de um mesmo líquido escoam separadamente, sem que ocorra mistura significativa entre as camadas, devido às baixas velocidades e à diferença de densidade (e temperatura). As consequências desse fenômeno não foram consideradas no projeto da maioria das centrais nucleares atualmente em operação. Entretanto, em alguns componentes de centrais nucleares, as diferenças de temperatura podem atingir cerca de 200 °C em uma região bastante estreita nas proximidades da interface entre as camadas de água fria e quente. Nesta condição, as fortes tensões geradas pelas diferenças de dilatação podem comprometer a integridade estrutural e a vida útil de tubulações relacionadas aos sistemas de segurança dessas centrais nucleares. Com o objetivo de estudar o fenômeno da estratificação foi projetada e construída a Instalação de Testes de Estratificação Térmica (ITET), tendo sido realizada uma série de experimentos simulando o bocal de injeção do gerador de vapor de uma central nuclear tipo PWR. Foram estudadas a evolução e as configurações de escoamento em regime de estratificação térmica, assim como a influência do número de Froude nos gradientes de temperatura, na posição da interface entre as camadas de água fria e de água quente e no aparecimento de oscilações desta interface. Os experimentos foram realizados com número de Froude variando de 0,02 a 0,4...Observação: O resumo, na íntegra, poderá ser visualizado no texto completo da tese digital / Abstract: One phase thermally stratified flows occur when two different layers of the same liquid at different temperatures flow separately in horizontal pipes without appreciable mixing due to the low velocities and difference in density (and temperature). The phenomenon was not considered in the design stage of most of the operating nuclear power plants. However, temperature differences of about 200 °C have been found in a narrow band around the hot and cold water interface in components under stratified flows. Loadings due to this phenomenon affected the integrity of safety related piping systems. The Thermal Stratification Test Facility (ITET), built to allow the experimental simulation of the thermal stratification, is presented so as the results of some experiments simulating one phase thermally stratified flows in geometry and flow condition similar to a nuclear reactor steam generator nozzle. They have the objective of studying the flow configurations and understanding the evolution of the of thermal stratification process. The driving parameter considered to characterize flow under stratified regime due to difference in specific masses is the Froude number. Different Froude numbers, from 0.02 to 0.4, were obtained in different testes by setting injection cold water flow rates and hot water initial temperatures as planned in the test matrix. Results are presented showing the influence of Froude number on the hot and cold water interface position, temperature gradients and striping phenomenon...Note: The complete abstract is available with the full electronic document / Doutorado / Sistemas de Processos Quimicos e Informatica / Doutor em Engenharia Química

Thesis. 1978. Nucl.E.--Massachusetts Institute of Technology. Dept. of Nuclear Engineering. / MICOFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / Nucl.E.

A fuel burnup model of a Pressurized Water Reactor (PWR) was developed based on one-group diffusion theory and used simple thermal cross sections. A computer program which simulates the depletion of the core of a PWR was written based on this model. The basic idea was to develop a fuel depletion program which could be readily understood by nuclear engineering students. Thus, accuracy was sacrificed for the sake of simplicity. The model was based upon a typical PWR with three concentric regions in the radial direction of differing fuel enrichment. Each of the regions was homogenized and the concentrations of the isotopes in each region were considered constant over a time interval. The isotopes considered were U-235, Pu-239, U-238, Xe-135, I-135, Sm-149, Pm-149 and the lumped burnable poison isotope. The flux was approximated by the sum of two trigonometric functions. The magnitude and shape of the flux were determined by holding power constant, constraining system to be critical and varying the soluble boron concentration to find the fla~test possible positive flux. A flux magnitude computed in this manner was compared to a similar flux magnitude given in a Final Safety Analysis Report. The concentrations of the isotopes were determined from the differential equations describing the rate of change of the concentrations. The behavior of the isotopes over core life was graphed and wherever possible compared to graphs from other sources. The concentrations calculated for U-235, U-238 and Pu-239 after 450 days were compared to the concentrations of the same isotopes calculated by a zero dimensional three-group model. The percentage difference between the concentrations determined by the two models varied from about 69% for Pu-239 to 1% for U-238. / Master of Science