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The world's growing need for safe, reliable, and long-term energy sources has intensified the research being conducted in the development of nuclear power. The operation of power reactors is contingent upon the continued availability of the fissile material required to maintain a critic a l reactor. The only naturally occuring fissile material is an isotope of uranium, U-235, which only accounts for 0.76 percent of the uranium that is mined. Due to the relative scarcity of this important fissile material it is estimated the United States' sources of economically recoverable fissile material will be delinquished with in about twenty-five years.^1

A new solution technique is derived for the time-dependent transport equation. This approach extends the steady-state coarse-mesh transport method that is based on global-local decompositions of large (i.e. full-core) neutron transport problems. The new method is based on polynomial expansions of the space, angle and time variables in a response-based formulation of the transport equation. The local problem (coarse mesh) solutions, which are entirely decoupled from each other, are characterized by space-, angle- and time-dependent response functions. These response functions are, in turn, used to couple an arbitrary sequence of local problems to form the solution of a much larger global problem. In the current work, the local problem (response function) computations are performed using the Monte Carlo method, while the global (coupling) problem is solved deterministically. The spatial coupling is performed by orthogonal polynomial expansions of the partial currents on the local problem surfaces, and similarly, the timedependent response of the system (i.e. the time-varying flux) is computed by convolving the time-dependent surface partial currents and time-dependent volumetric sources against pre-computed time-dependent response kernels.

Orientador: Prof. Dr. José Rubens Maiorino / Dissertação (mestrado) - Universidade Federal do ABC. Programa de Pós-Graduação em Energia, 2016. / O presente trabalho desenvolveu um código, intitulado STH-MOX-Th (Simplified Thermal- Hydraulics code-Mixed Oxide Thorium), com o objetivo de calcular os limites térmicos (temperaturas limite do combustível e do revestimento, além do DNBR-"Departure of Nucleate Boiling Ratio"- mínimo) de um reator PWR do tipo vareta para combustíveis de UO2 e óxidos mistos de Urânio-Tório. ((U,Th)O2) utilizando correlações específicas para cada combustível cujo coeficiente de condutividade térmica é uma função dependente da temperatura. Para tal resolução, foi utilizado o método de Runge-kutta de 4ª ordem. O código analisa apenas o canal mais quente do núcleo do reator e, por conta dessa simplificação, possui uma parte hidráulica simples. Além da parte hidráulica, o programa calcula a distribuição axial e radial das temperaturas do refrigerante e vareta, bem como distribuições de entalpia e pressão. Todos esses cálculos foram realizados no início do ciclo do combustível no caso do (U, Th)O2 e para o UO2 e, além disso, o código calcula casos considerando a queima do combustível (meio e final de ciclo) somente para o UO2, uma vez que não foi encontrada nenhuma correlação para o coeficiente de condutividade térmica para o (U,Th)O2 em função da queima. Para validar o programa foram utilizados dados referentes a usina de Angra 2 para a entrada do programa e os resultados comparados com os reportados pelo Relatório Final de Análise de Segurança da Eletronuclear e do reator AP-1000, desenvolvido pela Westinghouse. A grande contribuição do trabalho, é o cálculo dos limites térmicos de um reator utilizando óxidos mistos de urânio e tório no núcleo do reator AP-1000, que é objeto das pesquisas na UFABC. Apesar de não ser original, o trabalho possuí fins didáticos e será extremamente útil no que diz respeito a uma primeira análise dos limites térmicos de um reator nuclear. / The present study developed a code, named STH-MOX-Th (Simplified Thermal-Hydraulics code-Mixed Oxide Thorium), created in order to calculate the thermal limits (limit temperature of the fuel and of the coating, besides the DNBR -"Departure of Nucleate Boiling Ratio"- minimum) of a PWR rod type reactor to UO2 fuel and mixed oxides of Uranium- Thorium. ((U,Th)O2) using specific correlations to each fuel which coefficient of thermal conductivity is a function dependent on temperature. For such a resolution, the method Runge-kutta of 4th order was employed. The code analyses only the hottest channel of the reactor core and, because of this simplification, it has one simple hydraulic part. In addition to the hydraulic part, the program calculates the axial and radial distribution of refrigerant and rod temperatures, as well as the distributions of enthalpy and pressure. All these calculations were done in the beginning of the fuel cycle in the case of (U,Th)O2and, for UO2, the code also calculates cases that consider the fuel burning (beginning, middle and end of the fuel cycle) only for UO2, once it was not found any correlation to the coefficient of thermal conductivity to (U,Th)O2 being dependent on fuel burning.In order to validate the program, data from Angra 2 plant were used to the program input and the results were compared with the ones reported by the Final Report on Security Analysis of Eletronuclear and with the ones of AP-1000 reactor, developed by Westinghouse. As the main contribution, the program made such calculations to the project of the fuel reactor of (U-Th) O2, APTh-1000. Although this study is not original, it has learning purposes and will be extremely useful concerning a very first analysis of the thermal limits of a nuclear reactor.

Orientadores: Roger Josef Zemp, Paulo de Carvalho Tofani / Tese (doutorado) - Universidade Estadual de Campinas, Faculdade de Engenharia Quimica / Made available in DSpace on 2018-07-29T06:10:59Z (GMT). No. of bitstreams: 1 Navarro_MoysesAlberto_D.pdf: 5659553 bytes, checksum: 5053e728e8c53710ee79a2d3d789497c (MD5) Previous issue date: 2001 / Resumo: A limitação do escoamento em contracorrente, ou inundação, fenômeno caracterizado pelo controle que um gás exerce no escoamento de um líquido em sentido contrário, desempenha papel importante em diversos equipamentos das engenharias química e mecânica (condensador de refluxo, colunas recheio, tubos de calor etc.). Mais recentemente, o fenômeno tem recebido atenção especial da área nuclear devido à sua influência no comportamento termo fluido dinâmico de um reator nuclear durante um acidente de perda de refrigerante. A maioria dos estudos experimentais e analíticos sobre a inundação foi executada em tubos verticais. Menor atenção por parte dos pesquisadores tem recebido as geometrias mais complexas como as constituídas por canais de escoamento anulares, com placas perfuradas e, especialmente, aqueles constituídos de tubos horizontal e inclinado conectados por uma curva, como a "perna quente" dos reatores nucleares refrigerados a água pressurizada (Pressurized Water Reactor - PWR). Para melhor subsidiar as análises deste fenômeno, foi conduzido no CDTN/CNEN uma série de experimentos em seções de teste em acrílico com a mesma forma geométrica da "perna quente" de um PWR. Nestes experimentos, o escoamento em contracorrente foi estabelecido com injeção de água pela extremidade superior da tubulação inclinada e de ar através da outra extremidade da seção. Com incursão gradual ascendente na velocidade do ar para níveis de injeção de água preestabelecidos, foram determinadas as velocidades de início de arraste da água pelo ar, de início de arraste total e, durante a redução gradativa da velocidade do ar, as velocidades dos fluidos relativas à fase de inundação, quando a penetração da água é controlada pelo escoamento do ar. Com o objetivo de se avaliar as influências das características geométricas do canal de escoamento, foram realizados experimentos com diferentes comprimentos horizontais e inclinados, inclinações do duto inclinado, alturas hidrostáticas acima da extremidade superior e diâmetros da seção de testes. Os resultados experimentais mostraram uma dependência do inicio do arraste com a taxa de injeção água e com as características geométricas da seção de testes. Para uma geometria defmida, na condição de inundação, os pontos experimentais seguem uma curva característica até o início do arraste total assim como, na redução da vazão de ar, até o retomo à precipitação total, que independente da taxa de água injetada. Variações (:t 20°) em tomo de 50° na inclinação pouco afetam o comportamento da curva de inundação. Foi constatado ainda que, para uma mesma velocidade de ar, o aumento do comprimento horizontal ou do inclinado da seção provoca o aumento do arraste de água. O levantamento das influências dos parâmetros geométricos da seção de testes no comportamento da inundação gerou uma nova correlação para o fenômeno / Abstract: The Countercurrent Flow Limitation (CCFL), or flooding, is characterized by the restraint imposed by a gas on a countercurrent liquid flow. The phenomenon plays important role in several equipment in the chemical and mechanical engineerings (reflux condensator, packed columns, heat pipes). More recently the phenomenon has received special attention by the nuclear area due to its influence in the thermal-hydraulic behavior of a nuclear reactor during a postulated loss of coolant accident. Most of the experimental and analytical studies about the flooding was performed in vertical ducts. The more complex geometries, such as annular channels, or channels with perforated plates and, especially, those which are constituted by a horizontal pipe connected to an inclined riser, as the hot leg of the Pressurized Water Reactors - PWR, have received little attention by the researchers. To subsidize the analyses of this phenomenon, experiments in test sections with the same geometric form of the hot leg of a PWR were carried in CDTN/CNEN. In these experiments, the countercurrent flow was established with water injection in the upper extremity of the inclined pipe associated to air injection through the other extremity of the test section. With ascending and gradual air flow rate, for specific water flow rates, the air velocities at the onset of flooding and at the onset of total water carryover were measured. During an air flow rate reduction phase, the relative fluids velocities in the flooding phase, when the water penetration is controlled by the air in countercurrent, were also determined. In order to evaluate the influence of the geometric characteristics of the test section, experiments with different horizontal and inclined lengths, inclinations of the inclined riser, water head above the upper extremity and diameters of the test section, were also performed. The experimental results showed that the onset of flooding is a complex function of water flow late injection and depends on the geometry of the test section. For a specific geometry, in the flooding condition, the experimental points follow a characteristic curve ITom the onset of the total carryover until the total water precipitation. These flooding curve was found to be independent of the injected water flow rate. The imposed variations (:t 20) around 50° in inclination of the inclined riser produced negligible effects in the flooding curve. It was also verified that, for a same air velocity, a longer horizontal or inclined length induces an increase in the carryover water. This study proposes a new flooding correlation considering the influence of the geometrical parameters / Doutorado / Sistemas de Processos Quimicos e Informatica / Doutor em Engenharia Química

Thesis (M. S.)--Mechanical Engineering, Georgia Institute of Technology, 2008. / Committee Chair: Willem F.G. Van Rooijen; Committee Member: Ghiaasiaan, Seyed M; Committee Member: Weston M. Stacey

As the United States expands its quantity of nuclear reactors in the near future, the amount of spent nuclear fuel (SNF) will also increase. Closing the nuclear fuel cycle has become the next major technical challenge for the nuclear energy industry. By separating the transuranics (TRU) from the SNF discharged by Light Water Reactors, it is possible to fuel Advanced Burner Reactors to minimize the amount of SNF that must be stored in High Level Waste Repositories. One such ABR concept is the Subcritical Advanced Burner Reactor (SABR) being developed at the Georgia Institute of Technology. SABR is a subcritical, sodium-cooled fast reactor with a fusion neutron source capable of burning up to 25% of the TRU fuel over an 8.2 year residence time. In the SABR concept an annular core with a thickness of 0.6 m and an active height of 3.2 m surrounds the toroidal fusion neutron source. Neutron multiplication varies during the lifetime of the reactor from keff = 0.95 at the beginning of reactor life to 0.83 at the end of an equilibrium fuel cycle. Sixteen control rods worth 9$ are symmetrically positioned around the reactor. This thesis describes the dynamic safety analysis of the coupled neutron source, reactor core and reactor heat removal systems. A special purpose simulation model was written to predict steady-state conditions and accident scenarios in SABR by calculating the coupled evolution of the power output from the fusion and fission cores and the axial and radial temperature distributions of a fuel pin in the reactor. Reactivity Feedback was modeled for Doppler and sodium coolant voiding. SABR has a positive temperature reactivity feedback coefficient. A series of accident scenarios were simulated to determine how much time exists to implement corrective measures during an accident before damage to the reactor occurs.

Les Etudes Probabilistes de Sûreté (EPS) de niveau 2 en centrale nucléaire visent à identifier les séquences d’événements pouvant correspondre à la propagation d’un accident d’un endommagement du cœur jusqu’à une perte potentielle de l’intégrité de l’enceinte, et à estimer la fréquence d’apparition des différents scénarios possibles.<p>Ces accidents sévères dépendent non seulement de défaillances matérielles ou d’erreurs humaines, mais également de l’occurrence de phénomènes physiques, tels que des explosions vapeur ou hydrogène. La prise en compte de tels phénomènes dans le cadre booléen des arbres d’événements s’avère difficile, et les méthodologies dynamiques de réalisation des EPS sont censées fournir une manière plus cohérente d’intégrer l’évolution du processus physique dans les changements de configuration discrète de la centrale au long d’un transitoire accidentel.<p>Cette thèse décrit l’application d’une des plus récentes approches dynamiques des EPS – la Théorie de la Dynamique Probabiliste basée sur les Stimuli (SDTPD) – à différents modèles de déflagration d'hydrogène ainsi que les développements qui ont permis cette applications et les diverses améliorations et techniques qui ont été mises en oeuvre.<p><p>Level-2 Probabilistic Safety Analyses (PSA) of nuclear power plants aims to identify the possible sequences of events corresponding to an accident propagation from a core damage to a potential loss of integrity of the containment, and to assess the frequency of occurrence of the different scenarios.<p>These so-called severe accidents depend not only on hardware failures and human errors, but also on the occurrence of physical phenomena such as e.g. steam or hydrogen explosions. Handling these phenomena in the classical Boolean framework of event trees is not convenient, and dynamic methodologies to perform PSA studies are expected to provide a more consistent way of integrating the physical process evolution with the discrete changes of plant configuration along an accidental transient.<p>This PhD Thesis presents the application of one of the most recently proposed dynamic PSA methodologies, i.e. the Stimulus-Driven Theory of Probabilistic Dynamics (SDTPD), to several models of hydrogen explosion in the containment of a plant, as well as the developed methods and improvements.<p> / Doctorat en Sciences de l'ingénieur / info:eu-repo/semantics/nonPublished