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Hydrodynamic analysis of electron-beam heated UO₂ vaporization experimentsClark, Bradley Allan January 1979 (has links)
No description available.
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A health risk assessment for the decommissioning of the Georgia Institute of Technology Research reactorKiellman, Tracy Jo 12 1900 (has links)
No description available.
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An Improved Fission Product Pressure Model for Use in the Venus-II Disassembly CodeJensen, Ray Leland 01 April 1976 (has links)
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
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A coarse-mesh transport method for time-dependent reactor problemsPounders, Justin Michael 06 April 2010 (has links)
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.
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Estudo da limitação do escoamento em contracorrente agua/ar em canais horizontal e inclinado unidos por curvaNavarro, Moysés Alberto 20 December 2001 (has links)
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
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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
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A safety and dynamics analysis of the subcritical advanced burner reactor: SABRSumner, Tyler Scott January 2008 (has links)
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
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A safety and dynamics analysis of the subcritical advanced burner reactor: SABRSumner, Tyler Scott 03 June 2008 (has links)
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.
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Application of the Stimulus-Driven Theory of Probabilistic Dynamics to the hydrogen issue in level-2 PSA / Application de la Stimulus Driven Theory of Probabilistic Dynamics (SDTPD) au risque hydrogène dans les EPS de niveau 2.Peeters, Agnes 05 October 2007 (has links)
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
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