Algoritmos Numéricos para el Problema de Restauración de Imágenes usando el Método de las Proyecciones Alternantes

Directory of Open Access Journals (Sweden)

René Escalante

2011-04-01

Full Text Available Los algoritmos de proyección han evolucionado a partir del algoritmo de proyección alternante propuesto por J. von Neumann en 1933, donde el espacio solución es la intersección de un número finito de subespacios o conjuntos convexos. Investigaciones recientes se han centrado en técnicas para acelerar la convergencia del método y explotar el multiprocesamiento. En este trabajo consideramos el problema de restauración de imágenes. La mayoría de las técnicas desarrolladas para resolverlo han usado algoritmos iterativos; una de ellas consiste en usar proyecciones ortogonales alternantes. Llevamos a cabo una revisión cronológica de las diferentes técnicas en las que se ha aplicado el método de las proyecciones ortogonales alternantes al problema de restauración de imágenes, hasta llegar al enfoque reciente de Combettes (1997-1999, en donde la restauración se basa en el cálculo de proyecciones aproximadas, en lugar de proyecciones exactas.

Modelo origen destino para estimar el flujo de tráfico usando algoritmos genéticos

OpenAIRE

Aldás S., Milton R.; Flores C., Marco J.; Universidad de Cuenca; Dirección de Investigación de la Universidad de Cuenca; DIUC

2014-01-01

En este trabajo se ha desarrollado un nuevo método basado en Inteligencia Artificial para resolver un problema del matriz origen-destino (O-D) aplicado al caso de una red de tráfico vehicular en la ciudad de Ambato. El método implementado, basado en algoritmos genéticos (AG), resuelve el problema de minimización asociado al problema de matriz O-D. Para validar la técnica, se ha utilizado una red vial correspondiente a la zona del Mercado Modelo en la ciudad de Ambato, que es una zona de alta ...

BWR 90 and BWR 90+: Two advanced BWR design generations from ABB

International Nuclear Information System (INIS)

Haukeland, S.; Ivung, B.; Pedersen, T.

1999-01-01

ABB has two evolutionary advanced light water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and total power generation costs have been low. When developing the BWR 90 specific changes were introduced to a reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher than that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Hence, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The review work was completed in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with an 'evolutionary' design called BWR 90+, which aims at developing the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is performed by ABB Atom

Algoritmo de aceptación diferida matricial

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Jorge Oviedo

2008-01-01

Full Text Available En este artículo damos una versión matricial del algoritmo de aceptación diferida para el modelo de asignación (matching uno a uno. El algoritmo va modificando la matriz de preferencia de los agentes. Cuando el algoritmo se detiene se muestra que coincide con una asignación estable óptima de los agentes.

Algoritmo PSO para identificación de parámetros en un motor DC

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César Duarte

2010-01-01

Full Text Available En este artículo se presenta la aplicación del algoritmo Particle Swarm Optimization (PSO como estrategia de búsqueda para determinar los parámetros de un sistema tipo caja gris. Este proceso de identificación se ilustra utilizando la respuesta a un escalón de un motor DC en lazo abierto. Para agilizar la identificación del sistema a través del uso de PSO, se limita el espacio de búsqueda usando la información que puede extraerse de la respuesta en el dominio del tiempo del sistema a identificar. Los resultados del proceso de identificación de los parámetros del motor se obtienen utilizando la herramienta PSOt disponible para entorno Matlab® desarrollada por Brian Birge.

BWR 90: The ABB advanced BWR design

International Nuclear Information System (INIS)

Haukeland, S.; Ivung, B.; Pedersen, T.

1999-01-01

ABB has two evolutionary advanced fight water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and the total power generation costs have been low. In the development of BWR 90 specific changes were introduced to the reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher dim that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Thus, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The work is scheduled for completion in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with the 'evolutionary' design BWR 90+. The primary design goal is to develop the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is

Compact modular BWR (CM-BWR)

International Nuclear Information System (INIS)

Fennern, Larry; Boardman, Charles; Carroll, Douglas G.; Hida, Takahiko

2003-01-01

A preliminary assessment has shown that a small 350 MWe BWR reactor can be placed within a close fitting steel containment vessel that is 7.1 meters inside diameter. This allows the technology and manufacturing capability currently used to fabricate large ABWR reactor vessels to be used to provide a factory fabricated containment vessel for a 350 MWe BWR. When a close fitted steel containment is combined with a passive closed loop isolation condenser system and a natural circulating reactor system that contains a large water inventory, primary system leaks cannot uncover the core. This eliminates many of the safety systems needed in response to a LOCA that are common to large, conventional plant designs including. Emergency Core Flooding, Automatic Depressurization System, Active Residual Heat Removal, Safety Related Auxiliary Cooling, Safety Related Diesel Generators, Hydrogen Re-Combiners, Ex-vessel Core Retention and Cooling. By fabricating the containment in a factory and eliminating most of the conventional safety systems, the construction schedule is shortened and the capital cost reduced to levels that would not otherwise be possible for a relatively small modular BWR. This makes the CM-BWR a candidate for applications where smaller incremental power additions are desired relative to a large ALWR or where the local infrastructure is not able to accommodate a conventional ALWR plant rated at 1350 MWe or more. This paper presents a preliminary design description of a Compact Modular BWR (CM-BWR) whose design features dramatically reduce the size and cost of the reactor building and associated safety systems. (author)

Mapeo curvas típicas demanda de energía eléctrica del sector residencial, comercial e industrial de la ciudad de Medellín, usando redes neuronales artificiales y algoritmos de interpolación

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Héctor Tabares

2008-01-01

Full Text Available Uno de los principales problemas para modelar el consumo de energía eléctrica en un lugar determinado, consiste en la extracción del conocimiento cuando éste se encuentra almacenado en grandes volúmenes de información como, por ejemplo, registros históricos. De acuerdo con esta representación, cada hecho ocurrido y registrado está compuesto por una pareja de componentes (t, P en donde t representa el tiempo en el que se registro la muestra y P representa la potencia eléctrica consumida en ese instante. El registro diario cuenta con N casos que representa cada una de las parejas de estímulo-respuesta conocidas. El objetivo de este trabajo consiste en hallar una función que permita mapear el vector de variables de entrada t al vector de variables de salida P. donde F es una función cualquiera, en este caso el consumo de energía eléctrica. Su modelamiento con Redes Neuronales Artificiales (RNA es un Perceptron Multi Capa (PMC. Otra forma de modelarlo es usando Algoritmos de Interpolación (AI.

MODELACIÓN NUMÉRICA DE PROCESOS DE ESTERILIZACIÓN TÉRMICA DE ALIMENTOS USANDO VOLUMENES DE CONTROL: APROXIMACIÓN CILÍNDRICA

OpenAIRE

HÉCTOR J. CIRO-VELÁSQUEZ; CARLOS GONZÁLEZ; EDUARD GARCÍA

2009-01-01

Un proceso de esterilización térmica de alimentos fue modelado usando la técnica de diferenciación finita aplicada a volúmenes de control. El modelo de simulación fue desarrollado en coordenadas cilíndricas con un ejemplo de aplicación tomando como referencia un producto cárnico (carne de res). Los resultados de la simulación mostraron que el algoritmo desarrollado es independiente de la red nodal seleccionada, permitiendo cuantificar las curvas de penetración de calor en el punto crítico del...

Detección de daños en estructuras: estudio comparativo entre los algoritmos de optimización AGs y PSO

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Robinson Barrera

2014-01-01

Full Text Available En este artículo se comparan las técnicas de optimización de enjambre de partículas, PSO (de sus siglas en inglés, Particle Swarm Optimization y Algoritmos Genéticos (AG, ambas usadas en la implementación de un sistema para detectar y diagnosticar fallas en una estructura usando la respuesta modal. Diferentes episodios de daños (daño simple y daño múltiple con y sin ruido, permiten simular las condiciones reales en una estructura tipo viga y armadura para determinar el desempeño en la detección y diagnóstico de los sistemas propuestos. Adicionalmente, ambos sistemas propuestos se evalúan variando los volúmenes de información, es decir, el número de modos de vibración. La respuesta dinámica de la estructura en condición normal o de falla se obtiene usando la herramienta de uso libre OPENSEES® y los algoritmos de optimización PSO y AGs se implementan en ambiente Matlab®. La comparación de desempeño de las dos técnicas en cuanto a sus capacidades de detección y diagnóstico se realizan en una viga segmentada en 10 partes y simplemente apoyada y una armadura de 13 elementos. Resultados experimentales mostraron la eficacia y robustez de los sistemas propuestos en la determinación del estado del sistema en diversos ambientes de ruido y con diferentes volúmenes de información. Sin embargo su desempeño varía según el sistema monitoreado y la función objetivo utilizada

Algoritmo de alocação de recursos discretos com análise de envoltória de dados

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João Carlos Correia Baptista Soares de Mello

2006-08-01

Full Text Available A alocação de recursos é um dos problemas clássicos abordados pela Pesquisa Operacional. A inserção de modelos de Análise de Envoltória de Dados (DEA nesse contexto traz novas possibilidades de estudo, em especial quando combinada com o uso de algoritmos de programação inteira. Este artigo apresenta um algoritmo para distribuição de recursos discretos usando modelos DEA de forma seqüencial. O método proposto é aplicado no estudo de caso da alocação de vagas docentes na Universidade Federal Fluminense. Os resultados são comparados com aqueles obtidos pela comissão encarregada da distribuição das vagas.The resource allocation is one of the main problems in Operational Research. The use of Data Envelopment Analysis (DEA in this field is a new feature with a great potential, mainly when combined with integer programming problems. This paper presents an algorithm to allocate integer resources using a step-by-step DEA algorithm. We applied the proposed approach to a real case study, which consists in allocating teacher positions in some departments of Universidade Federal Fluminense. We compare the results with those obtained by the official commission.

Estimativa do balanço de energia utilizando imagens TM - Landsat 5 e o algoritmo SEBAL no litoral sul de Pernambuco

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Célia Cristina Machado

2014-03-01

Full Text Available As trocas de energia na interface solo-planta-atmosfera, por meio das componentes do balanço de radiação (Rn e dos fluxos de calor no solo (G, sensível (H e latente (LE, são essenciais para a modelagem climática e hidrológica que, por sua vez, afetam toda a biosfera. Dessa forma, esta pesquisa teve por objetivos: (1 estimar e comparar o comportamento dos componentes do balanço de energia, usando o algoritmo SEBAL - Surface Energy Balance Algorithm for Land, em diferentes tipos de uso e cobertura do solo e (2 validar o algoritmo. Para isso, foram usadas três imagens TM - Landsat 5 de órbita e ponto 214/066, com datas de passagem em 26 de agosto de 2006, 6 de setembro de 2010 e 28 de janeiro de 2011, todas recortadas para enquadrar o litoral sul de Pernambuco. Os resultados mostraram a capacidade do SEBAL de estimar a variabilidade espacial dos diferentes componentes do balanço de energia, com capacidade de distinguir diferentes ocupações do solo. O cômputo do balanço de energia possibilita observar as formas diferenciadas do uso da energia pelos diferentes tipos de cobertura do solo e verificar aqueles que propiciam melhor conforto térmico. As estimativas da evapotranspiração obtidas pelo algoritmo SEBAL apresentaram uma boa concordância com os resultados obtidos pelo método da FAO-Penman-Monteith demonstrando, dessa forma, a potencialidade da abordagem metodológica escolhida.

Los algoritmos como instrumento de la matemática

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Luis Roberto Ojeda Ch.

1993-07-01

Full Text Available El concepto de algoritmo es fundamental en múltiples campos de la actividad humana. Se relacionan en este artfculo los conceptos de función y de algoritmo para luego caracterizar un tipo especial de función: el de las recursivas. Finalmente se establece la Tesis de Alonzo Church que indica que sólo para las funciones recursivas pueden obtenerse algoritmos de evaluación.

Algoritmos de antialiasing

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Gabriel Mañana Guichón

1993-07-01

Full Text Available Se presentan los algoritmos desarrollados para el trazado de lineas suaves y aplicación de texto sobre imágenes, como ejemplo del trabajo de investigación que se lleva a cabo en el campo del filtrado de imágenes o antialiasing.

A intensidade dos algoritmos nas séries iniciais: uma imposição sócio-histórico-estrutural ou opção valiosa? p.55-76

OpenAIRE

Mendonça, Maria do Carmo Domite

2009-01-01

Como nós professores podemos reconhecer/valorizar ou não o quadro que aí está em Educação Matemática? Como os professores-pesquisadores podemos compreender a ansiedade existente em treinar os algoritmos convencionais, quando o assunto é o cálculo das operações aritméticas básicas? Um caminho para a mudança seria focalizar/discutir o que, naturalmente, tem bloqueado atitudes de maior autonomia frente a esse fato matemático, imobilizando-os mesmo para qualquer inovação. Então, usando argumentos...

Estudo comparativo entre algoritmos das transformadas discretas de Fourier e Wavelet

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Wilson Hissamu Shirado

2015-11-01

Full Text Available Este trabalho apresenta um estudo comparativo das complexidades dos algoritmos das Transformadas Discretas de Fourier, Wavelet e Transformada Rápida de Fourier. As formalizações matemáticas e algumas características dos algoritmos são apresentadas, assim como alguns conceitos de complexidade assintótica. Por fim, é realizado um ensaio prático para comparação dos algoritmos, abrangendo questões como tempo de execução, vantagens e desvantagens de cada transformada assim como avaliações a respeito das diferentes resoluções tempo/frequência de cada algoritmo.

Framework de desarrollo rápido de algoritmos paralelos

OpenAIRE

Expósito Piñol, Luis

2012-01-01

El proyecto ofrece un framework de programación de algoritmos paralelos basado en el algoritmo genético BRKGA, permitiendo resolver diferentes problemas solo cambiando una parte del framework. Todo esto utilizando la GPU para tener una mayor potencia de cálculo.

¿Podemos predecir en algoritmos paralelos no deterministas?

OpenAIRE

Fritzsche, Paula Cecilia

2007-01-01

¿Podemos predecir en algoritmos paralelos no deterministas? Sí, es posible. A lo largo de este documento se demostrará esta afirmación. La evaluación científica de algoritmos para solucionar todo tipo de problemas es uno de los puntos clave en ciencias de la computación. En el terreno de ciencia computacional (una disciplina emergente) continuamente surgen nuevos desafíos.

Usando algoritmos e ábaco no estudo do sistema de numeração decimal em um curso de Pedagogia

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Viegas, Elis Regina dos Santos

2015-05-01

Full Text Available Studies in Mathematics Education have described difficulties of teachers in the use of algorithms in arithmetic teaching in the initial grades, difficulties leading to questions about the quality of theoretical and practical training provided for teachers in this area. In this context, the present study aimed to investigate the formation of concepts and construction of mathematical knowledge of Pedagogy students in relation to arithmetic teaching, basic operations using abacus and algorithms, as well as the decimal numbering system. In the theoretical approach, the authors reflect on initial teacher training for Mathematics teachers in the initial grades, the necessity of training grounded in knowledge to the process of teaching/learning and the importance of teacher training in methodological contents and in specific contents. A qualitative and quantitative approach was used, the data collection occurred with the use of semi-structured questionnaires, procedures in the activities report, direct observation and video recording. The results showed that studies with both methodological approaches, abacus and algorithms, might contribute to the formation, reframing concepts and construction of mathematical knowledge, but also could provide information for a better pedagogical and specific training. Estudos em Educação Matemática têm descrito dificuldades de professores na utilização de algoritmos no ensino de aritmética nas séries iniciais, dificuldades que provocam questionamentos quanto à qualidade da formação teórico-prática oferecida para a docência nessa área. Nesse contexto, o presente estudo teve como objetivo investigar a formação de conceitos e construção do conhecimento matemático de estudantes de Pedagogia em relação ao ensino de aritmética, das operações fundamentais com a utilização de ábaco e algoritmos, bem como, do sistema de numeração decimal. No referencial teórico adotado, os autores refletem sobre

Development of a BWR core burn-up calculation code COREBN-BWR

International Nuclear Information System (INIS)

Morimoto, Yuichi; Okumura, Keisuke

1992-05-01

In order to evaluate core performances of BWR type reactors, the three dimensional core burnup calculation code COREBN-BWR and the fuel management code HIST-BWR have been developed. In analyses of BWR type reactors, thermal hydraulics calculations must be coupled with neutronics calculations to evaluate core performances, because steam void distribution changes according to the change of the power distribution. By installing new functions as follows to the three dimensional core burnup code COREBN2 developed in JAERI for PWR type reactor analyses, the code system becomes to be applicable to burnup analyses of BWR type reactors. (1) Macroscopic cross section calculation function taking into account of coolant void distribution. (2) Thermal hydraulics calculation function to evaluate core flow split, coolant void distribution and thermal margin. (3) Burnup calculation function under the Haling strategy. (4) Fuel management function to incorporate the thermal hydraulics information. This report consists of the general description, calculational models, input data requirements and their explanations, detailed information on usage and sample input. (author)

Algoritmos para la clasificación piramidal simbólica

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Oldemar Rodríguez

2009-02-01

Full Text Available En este artículo se define el concepto de pirámide simbólica, además se presentan dos algoritmos para generar este tipo d pirámide a partir de una matriz de datos simbólicos. El primer algoritmo (CAPS encuentra un "orden total compatible con la pirámide" de los n objetos, mientras que el segundo (CAPSO construye la pirámide a partir de un orden dado apriori en los objetos, dicho orden se recibe como entrada en el algoritmo. Ambos algoritmos, además de producir la pirámide, para cada grada encuentran el objeto simbólico asociado a cada nodo y su extensión. También se presentan los teoremas de convergencia. Palabras clave: pirámide, objeto simbólico, grada, grado de generalidad, objeto completo, componente conexa, tablas de datos simbólica.

Estudo comparativo entre algoritmos das transformadas discretas de Fourier e Wavelet

OpenAIRE

Wilson Hissamu Shirado; Márcio de Abreu Moreira; Jandira Guenka Palma; Sylvio Barbon Júnior

2015-01-01

Este trabalho apresenta um estudo comparativo das complexidades dos algoritmos das Transformadas Discretas de Fourier, Wavelet e Transformada Rápida de Fourier. As formalizações matemáticas e algumas características dos algoritmos são apresentadas, assim como alguns conceitos de complexidade assintótica. Por fim, é realizado um ensaio prático para comparação dos algoritmos, abrangendo questões como tempo de execução, vantagens e desvantagens de cada transformada assim como avaliações a respei...

Algoritmo de cifrado para sistemas móviles

OpenAIRE

Castro Lechtaler, Antonio; Cipriano, Marcelo; García, Edith; Liporace, Julio César; Maiorano, Ariel; Malvacio, Eduardo; Tapia, Néstor

2017-01-01

Este proyecto de investigación persigue elaborar el diseño y desarrollo de un Algoritmo de Cifrado para Sistemas Móviles que por sus propiedades de velocidad, compactibili- dad y robustez; pueda ser implementado en equipos de comunicaciones que funcionan so- bre Sistemas Móviles. Se esperan obtener resultados teóricos, prácticos y la realización de un desarrollo experimental. Los modernos algoritmos criptológicos responden a principios y filosofías diferentes a las que se llevaban a cab...

NUEVO ALGORITMO MULTICLASIFICADOR PARA FLUJOS DE DATOS CON CAMBIOS DE CONCEPTO

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Ramón Osmany Ramírez Tasé

2016-04-01

Full Text Available Los algoritmos multiclasificadores se han mostrado particularmente eficientes para trabajar sobre espacios de datos grandes y complejos como los llamados flujos de datos. En estos flujos, durante la clasificación, aparecen conceptos que cambian con el tiempo, por lo que los métodos para su minería, sobre todo los  que detectan y se adaptan  a estos cambios, son importantes por su aplicación en áreas como: bioinformática, medicina, economía y finanzas, industria, medio ambiente, entre otras. La presente investigación propone un nuevo algoritmo multiclasificador que se adapta a los cambios de conceptos, tiene votación ponderada con una nueva forma para ajustar los pesos y permite variar el tipo de clasificador básico. El algoritmo fue implementado en compatibilidad y bajo las exigencias del entorno de trabajo MOA (Massive Online Analysis facilitando la comparación con otros algoritmos conocidos y la generación de bases de datos sintéticas que simulan cambios de conceptos. Para la experimentación se generaron experiencias bajo conceptos artificiales conocidos, tales como: SEA, LED, STAGGER e Hiperplano; logrando mostrar la alta capacidad de adaptación y la estabilidad del algoritmo frente a diferentes situaciones simuladas.

PLANEAMIENTO MULTIOBJETIVO DE SISTEMAS DE DISTRIBUCIÓN USANDO UN ALGORITMO EVOLUTIVO NSGA-II PLANEJAMENTO MULTIOBJETIVO DE SISTEMAS DE DISTRIBUIÇÃO USANDO UM ALGORITMO EVOLUTIVO NSGA-II MULTIOBJECTIVE DISTRIBUTION SYSTEM PLANNING USING AN NSGA-II EVOLUTIONARY ALGORITHM

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Libardo López

2011-06-01

Full Text Available En este artículo se presenta una metodología para solucionar el problema del planeamiento de sistemas de distribución empleando una técnica de optimización multiobjetivo. En el modelo propuesto se consideran la ubicación y dimensionamiento de nuevos elementos y la repotenciación de elementos existentes. En el planteamiento del problema se incluyen en la función objetivo los costos de inversión y de operación y la confiabilidad de la red. Este tema de investigación es de relevancia para las empresas distribuidoras de energía eléctrica, ya que permite contar con nuevas herramientas que acercan el problema a situaciones reales, tales como considerar varios objetivos, lo cual incide favorablemente en las finanzas y en la operación del sistema. En la solución del modelo se emplea un algoritmo elitista de ordenamiento no dominado (NSGA-II, y para verificar su eficiencia se recurre a un caso de la literatura especializada, que corresponde a una red de distribución de un sistema eléctrico.Neste artigo apresenta-se uma metodologia para solucionar o problema do planejamento de sistemas de distribuição empregando uma técnica de otimização multiobjetivo. No modelo proposto considera-se a localização e dimensionamento de novos elementos e a repotenciação de elementos existentes. Na abordagem do problema incluem-se na função objetivo os custos de investimento e de operação e a confiabilidade da rede. Este tema de pesquisa é de relevância para as empresas distribuidoras de energia elétrica, já que permite contar com novas ferramentas que acercam o problema a situações reais, tais como considerar vários objetivos, o qual incide favoravelmente nas finanças e na operação do sistema. Na solução do modelo emprega-se um algoritmo elitista de ordenamento não-dominado (NSGA-II, e para verificar sua eficiência se emprega um caso da literatura especializada, que corresponde a uma rede de distribuição de um sistema el

Una comparación de algoritmos evolutivos para la optimización de funciones multimodales

OpenAIRE

Brero, Alejandro C.; Gallard, Raúl Hector

2000-01-01

La Computación Evolutiva (CE) ha sido reconocida recientemente como un campo de investigación que estudia un nuevo tipo de algoritmos: los algoritmos evolutivos (AEs). Opuestamente a los enfoques tradicionales que mejoran una única solución, estos algoritmos procesan poblaciones de soluciones y poseen como características comunes la reproducción, la variación aleatoria, la competición y la selección de individuos. Los algoritmos genéticos (AGs) y las estrategias evolutivas (EEs) constituye...

Algoritmos multirecombinativos aplicados al problema de ruteo de vehículos

Directory of Open Access Journals (Sweden)

Renzo Ramon Miño

2014-06-01

Full Text Available El diseño de algoritmos eficientes para resolver problemas complejos ha sido tradicionalmente uno de los aspectos más importantes en la investigación en el campo de la informática. El objetivo perseguido en este campo es fundamentalmente el desarrollo de nuevos métodos capaces de resolver problemas complejos con el menor esfuerzo computacional posible, mejorando así a los algoritmos existentes. En consecuencia, esto no sólo permite afrontar los problemas de forma más eficiente, sino afrontar tareas vedadas en el pasado debido a su alto costo computacional. Las metaheurísticas son métodos que integran procedimientos de mejora local y estrategias de alto nivel para realizar una búsqueda robusta en el espacio-problema. El problema de ruteo de vehículos es un problema de optimización combinatoria de gran importancia en diferentes entornos logísticos debido a su dificultad (NP-duros y a sus múltiples aplicaciones industriales. Se han propuestos varias soluciones a este problema haciendo uso de heurísticas y metaheurísticas. En este trabajo proponemos dos algoritmos para resolver el problema de ruteo de vehículos con capacidad limitada, utilizando como base un Algoritmo Evolutivo conocido como MCMP-SRI (Stud and Random Inmigrants combinados con Hill- Climbing. Detalles de los algoritmos y los resultados de los experimentos muestran un promisorio comportamiento para resolver el problema.

Solución de un problema de secuenciamiento en configuración jobshop flexible utilizando un Algoritmo de Estimación de Distribuciones.

Directory of Open Access Journals (Sweden)

R. Pérez

2015-01-01

Full Text Available Resumen: El problema de secuenciamiento en configuración jobshop flexible se refiere en asignar operaciones de diversos trabajos a un conjunto de equipos disponibles. La asignación mencionada busca generalmente minimizar el tiempo de culminación de todos los trabajos conocido en la literatura como ‘makespan’. Se propone un Algoritmo de Estimación de Distribuciones para Secuenciamiento, AEDS por simplicidad y funcionalidad. El AEDS intenta encontrar una relación o interacción entre las variables de entrada, trabajos, operaciones y turnos de trabajo, para optimizar la variable de salida de procesos de manufactura reales, el makespan. En este sentido el algoritmo AEDS es usado para guiar la búsqueda y resolver el problema. En el marco del algoritmo, tres modelos gráficos fueron usados para encontrar mejores soluciones. Establecer horas fuera de servicio para los operadores antes de iniciar sus actividades en cada turno de trabajo como parámetro de entrada y su desarrollo a través del algoritmo AEDS constituye una novedad de esta investigación respecto los trabajos existentes. La comparación entre AEDS y un algoritmo genético muestra la efectividad del AEDS resolviendo el problema enunciado. Usando el AEDS propuesto, el desempeño de procesos de manufactura reales mejora notablemente cuando diferentes equipos disponibles son asignados a diferentes programas de producción. Abstract: The flexible jobshop scheduling problem permits the operation of each job to be processed by more than one machine. The configuration mentioned generally seeks to minimize the completion time of all jobs known in the literature as ‘makespan’. We propose an Estimation of Distribution Algorithm for Sequencing, AEDS for simplicity and functionality. The AEDS attempts to find a relationship or interaction between the input variables, jobs, operations and shifts to optimize the output variable of real

Obtention control bars patterns for a BWR using Tabo search; Obtencion de patrones de barras de control para un BWR usando busqueda Tabu

Energy Technology Data Exchange (ETDEWEB)

Castillo, A.; Ortiz, J.J.; Alonso, G. [Instituto Nacional de Investigaciones Nucleares, Km. 36.5 Carretera Mexico-Toluca, Ocoyoacac, Estado de Mexico 52045 (Mexico); Morales, L.B. [UNAM, IIMAS, Ciudad Universitaria, D. F. 04510 (Mexico); Valle, E. del [IPN, ESFM, Unidad Profesional ' Adolfo Lopez Mateos' , Col. Lindavista 07738, D. F. (Mexico)]. e-mail: jacm@nuclear.inin.mx

2004-07-01

The obtained results when implementing the technique of tabu search, for to optimize patterns of control bars in a BWR type reactor, using the CM-PRESTO code are presented. The patterns of control bars were obtained for the designs of fuel reloads obtained in a previous work, using the same technique. The obtained results correspond to a cycle of 18 months using 112 fresh fuels enriched at the 3.53 of U-235. The used technique of tabu search, prohibits recently visited movements, in the position that correspond to the axial positions of the control bars, additionally the tiempo{sub t}abu matrix is used for to manage a size of variable tabu list and the objective function is punished with the frequency of the forbidden movements. The obtained patterns of control bars improve the longitude of the cycle with regard to the reference values and they complete the restrictions of safety. (Author)

The BWR Stability Issue

International Nuclear Information System (INIS)

D'Auria, F.

2008-01-01

The purpose of this paper is to supply general information about Boiling Water Reactor (BWR) stability. The main concerned topics are: phenomenological aspects, experimental database, modelling features and capabilities, numerical models, three-dimensional modelling, BWR system performance during stability, stability monitoring and licensing aspects.

Modelación númerica de procesos de esterilización térmica de alimentos usando volúmenes de control: Aproximación cilíndrica./Numerical simulation of thermal sterilization process of foods using control volumes: Cylindrical aproximation.

OpenAIRE

Ciro Velásquez, Héctor José; González, Carlos; García, Eduardo

2009-01-01

Un proceso de esterilización térmica de alimentos fue modelado usando la técnica de diferenciación finita aplicada a volúmenes de control. El modelo de simulación fue desarrollado en coordenadas cilíndricas con un ejemplo de aplicación tomando como referencia un producto cárnico (carne de res). Los resultados de la simulación mostraron que el algoritmo desarrollado es independiente de la red nodal seleccionada, permitiendo cuantificar las curvas de penetración de calor en el punto crítico del...

SCORPIO-BWR: status and future plans

International Nuclear Information System (INIS)

Porsmyr, Jan; Bodal, Terje; Beere, William H.

2004-01-01

Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR system on a

SCORPIO-BWR: status and future plans

Energy Technology Data Exchange (ETDEWEB)

Porsmyr, Jan; Bodal, Terje; Beere, William H. (and others)

2004-07-01

Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR

Identification of related multilingual documents using ant clustering algorithms Identificación de documentos multilingües relacionados mediante algoritmos de clustering de hormigas

Directory of Open Access Journals (Sweden)

Ángel Cobo

2011-12-01

Full Text Available This paper presents a document representation strategy and a bio-inspired algorithm to cluster multilingual collections of documents in the field of economics and business. The proposed approach allows the user to identify groups of related economics documents written in Spanish and English using techniques inspired on clustering and sorting behaviours observed in some types of ants. In order to obtain a language independent vector representation of each document two multilingual resources are used: an economic glossary and a thesaurus. Each document is represented using four feature vectors: words, proper names, economic terms in the glossary and thesaurus descriptors. The proper name identification, word extraction and lemmatization are performed using specific tools. The tf-idf scheme is used to measure the importance of each feature in the document, and a convex linear combination of angular separations between feature vectors is used as similarity measure of documents. The paper shows experimental results of the application of the proposed algorithm in a Spanish-English corpus of research papers in economics and management areas. The results demonstrate the usefulness and effectiveness of the ant clustering algorithm and the proposed representation scheme.Este artículo presenta una estrategia de representación documental y un algoritmo bioinspirado para realizar procesos de agrupamiento en colecciones multilingües de documentos en las áreas de la economía y la empresa. El enfoque propuesto permite al usuario identificar grupos de documentos económicos relacionados escritos en español o inglés usando técnicas inspiradas en comportamientos de organización y agrupamiento de objetos observados en algunos tipos de hormigas. Para conseguir una representación vectorial de cada documento independiente del idioma, se han utilizado dos recursos lingüísticos: un glosario económico y un tesauro. Cada documento es representado usando

High Fidelity BWR Fuel Simulations

Energy Technology Data Exchange (ETDEWEB)

Yoon, Su Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-08-01

This report describes the Consortium for Advanced Simulation of Light Water Reactors (CASL) work conducted for completion of the Thermal Hydraulics Methods (THM) Level 3 milestone THM.CFD.P13.03: High Fidelity BWR Fuel Simulation. High fidelity computational fluid dynamics (CFD) simulation for Boiling Water Reactor (BWR) was conducted to investigate the applicability and robustness performance of BWR closures. As a preliminary study, a CFD model with simplified Ferrule spacer grid geometry of NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) benchmark has been implemented. Performance of multiphase segregated solver with baseline boiling closures has been evaluated. Although the mean values of void fraction and exit quality of CFD result for BFBT case 4101-61 agreed with experimental data, the local void distribution was not predicted accurately. The mesh quality was one of the critical factors to obtain converged result. The stability and robustness of the simulation was mainly affected by the mesh quality, combination of BWR closure models. In addition, the CFD modeling of fully-detailed spacer grid geometry with mixing vane is necessary for improving the accuracy of CFD simulation.

BWR core melt progression phenomena: Experimental analyses

International Nuclear Information System (INIS)

Ott, L.J.

1992-01-01

In the BWR Core Melt in Progression Phenomena Program, experimental results concerning severe fuel damage and core melt progression in BWR core geometry are used to evaluate existing models of the governing phenomena. These include control blade eutectic liquefaction and the subsequent relocation and attack on the channel box structure; oxidation heating and hydrogen generation; Zircaloy melting and relocation; and the continuing oxidation of zirconium with metallic blockage formation. Integral data have been obtained from the BWR DF-4 experiment in the ACRR and from BWR tests in the German CORA exreactor fuel-damage test facility. Additional integral data will be obtained from new CORA BWR test, the full-length FLHT-6 BWR test in the NRU test reactor, and the new program of exreactor experiments at Sandia National Laboratories (SNL) on metallic melt relocation and blockage formation. an essential part of this activity is interpretation and use of the results of the BWR tests. The Oak Ridge National Laboratory (ORNL) has developed experiment-specific models for analysis of the BWR experiments; to date, these models have permitted far more precise analyses of the conditions in these experiments than has previously been available. These analyses have provided a basis for more accurate interpretation of the phenomena that the experiments are intended to investigate. The results of posttest analyses of BWR experiments are discussed and significant findings from these analyses are explained. The ORNL control blade/canister models with materials interaction, relocation and blockage models are currently being implemented in SCDAP/RELAP5 as an optional structural component

Optimization of fuel reloads for a BWR using the ant colony system; Optimizacion de recargas de combustible para un BWR usando el sistema de colonia de hormigas

Energy Technology Data Exchange (ETDEWEB)

Esquivel E, J. [Universidad Autonoma del Estado de Mexico, Facultad de Ingenieria, Cerro de Coatepec s/n, Ciudad Universitaria, 50110 Toluca, Estado de Mexico (Mexico); Ortiz S, J. J. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: jaime.es.jaime@gmail.com

2009-10-15

In this work some results obtained during the development of optimization systems are presented, which are employees for the fuel reload design in a BWR. The systems use the ant colony optimization technique. As first instance, a system is developed that was adapted at travel salesman problem applied for the 32 state capitals of Mexican Republic. The purpose of this implementation is that a similarity exists with the design of fuel reload, since the two problems are of combinatorial optimization with decision variables that have similarity between both. The system was coupled to simulator SIMULATE-3, obtaining good results when being applied to an operation cycle in equilibrium for reactors of nuclear power plant of Laguna Verde. (Author)

ESTIMACIÓN DE PARÁMETROS EN MODELOS NO LINEALES: ALGORITMOS Y APLICACIONES

Directory of Open Access Journals (Sweden)

Óscar Cornejo Zúñiga

Full Text Available En este artículo se muestran diferentes algoritmos para estimar parámetros en modelos no lineales. Se aplican primeramente a una base de datos de problemas clasificados difíciles. Posteriormente, se muestra el comportamiento de los algoritmos para el estudio de crecimiento de la merluza común en machos y hembras, anchoveta y sardina común ajustando un modelo de Von Bertalanffy. Se aplica el test de Cerrato para la comparación de crecimientos entre géneros para la merluza común. Los algoritmos se implementaron en ambiente MATLAB presentando un buen comportamiento en cuanto a tiempo CPU, número de iteraciones y exactitud de la solución encontrada respecto de valores certificados de los problemas de la base de datos.

Development of advanced BWR

International Nuclear Information System (INIS)

Toyota, Masatoshi

1982-01-01

The Japanese technology and domestic production of BWR type nuclear power plants have been established through the experiences in the construction and operation of BWRs in addition to the technical agreement with the General Electric Co. In early days, the plants experienced some trouble such as stress corrosion cracking and some inconvenience in the operation and maintenance. The government, electric power companies and BWR manufacturers have endeavored to standardize and improve the design of LWRs for the purpose of improving the safety, reliability and the rate of operation and reducing the radiation exposure dose of plant workers. The first and second stages of the standardization and improvement of LWRs have been completed. Five manufacturers of BWRs in the world have continued the conceptual design of a new version of BWR power plants. It was concluded that this is the most desirable version of BWR nuclear power stations, but the technical and economic evaluation must be made before the commercial application. Six electric power companies and three manufacturers of BWRs in Japan set up the organization to develop the technology in cooperation. The internal pump system, the new control rod drive mechanism and others are the main features. (Kako, I.)

ANÁLISIS Y COMPARACIÓN DE LOS ALGORITMOS DE CIFRADOS SIMÉTRICO EN PLATAFORMAS WINDIWS

OpenAIRE

Carrera Albán, Thruman Wladimir

2016-01-01

El objetivo principal que guía el diseño de cualquier algoritmo de cifrado debe ser la seguridad contra ataques no autorizados. Sin embargo, para todas las aplicaciones prácticas, el rendimiento y la velocidad también son preocupaciones importantes. Estas son las dos características principales que diferencian un algoritmo de cifrado de otro. Este documento proporciona la comparación de rendimiento entre dos de los algoritmos de cifrado más utilizados: AES (Rijndael) y Blowfish. La comparació...

Extensión del Algoritmo Evolutivo MOS con técnicas de Evolución Diferencial

OpenAIRE

Fernández Martín, Jaime

2008-01-01

Integración de un nuevo algoritmo de Evolución Diferencial en una librería de Algoritmos Evolutivos y su adaptación en forma de técnica para el algoritmo de hibridación MOS (Multiple Offspring Sampling). MOS es capaz de adaptarse a los problemas, premiando a las técnicas que mejores resutlados vayan obteniendo. Pruebas y estudios sobre los resultados obtenidos con esta nueva técnica.

Aplicando un Algoritmo Genético para Balancear Carga Dinámicamente en Ambientes Distribuidos Orientados a Objetos (CORBA)

OpenAIRE

Fco. Javier Luna Rosas; Rene Tristán Ávila; J. de Jesús Martínez Pedroza

2003-01-01

Balancear Carga significa como distribuir procesos entre procesadores conectados por una red, para equilibrar la carga de trabajo entre ellos. Los algoritmos de planeación distribuida global pueden ser divididos en dos grandes grupos: algoritmos de balanceo de carga dinámica y algoritmos de balanceo de carga estática. Los algoritmos de balanceo de carga estática, también referenciados como planeación de tareas obtienen la localización de todos sus requerimientos antes de comenzar su ...

BWR Radiation Assessment and Control Program: assessment and control of BWR radiation fields. Volume 1. Executive summary

International Nuclear Information System (INIS)

Anstine, L.D.

1983-05-01

This report covers work on the BWR Radiation Assessment and Control (BRAC) Program from 1978 to 1982. The major activities during this report period were assessment of the radiation-level trends in BWRs, evaluation of the effects of forward-pumped heater drains on BWR water quality, installation and operation of a corrosion-product deposition loop in an operating BWR, and analyzation of fuel-deposit samples from two BWRs. Radiation fields were found to be controlled by cobalt-60 and to vary from as low as 50 mr/hr to as high as 800 mr/hr on the recirculation-system piping. Detailed information on BWR corrosion films and system deposits is presented in the report. Additionally, the results of an oxygen-injection experiment and recontamination monitoring studies are provided

ANÁLISIS A LOS ALGORITMOS UTILIZADOS POR EL SOFTWARE PARA BI

Directory of Open Access Journals (Sweden)

CURTIS COOKE

2011-12-01

Full Text Available El propósito de este trabajo es examinar cuáles son los algoritmos que utilizan los paquetes de software más populares para BI. Una encuesta inicial informal muestra que los actuales paquetes de software de BI incluyen a la mayoría de algoritmos descritos en la literatura de los paquetes estadísticos tradicionales. Además, una de las principales diferencias entre el software para BI y los paquetes estadísticos tradicionales es que el primero tiene mejores características para los reportes.

BWR Full Integral Simulation Test (FIST) Phase II test results and TRAC-BWR model qualification

International Nuclear Information System (INIS)

Sutherland, W.A.; Alamgir, M.; Findlay, J.A.; Hwang, W.S.

1985-10-01

Eight matrix tests were conducted in the FIST Phase I. These tests investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. There are nine tests in Phase II of the FIST program. They include the following LOCA tests: BWR/6 LPCI line break, BWR/6 intermediate size recirculation break, and a BWR/4 large break. Steady state natural circulation tests with feedwater makeup performed at high and low pressure, and at high pressure with HPCS makeup, are included. Simulation of a transient without rod insertion, and with controlled depressurization, was performed. Also included is a simulation of the Peach Bottom turbine trip test. The final two tests simulated a failure to maintain water level during a postulated accident. A FIST program objective is to assess the TRAC code by comparisons with test data. Two post-test predictions made with TRACB04 are compared with Phase II test data in this report. These are for the BWR/6 LPCI line break LOCA, and the Peach Bottom turbine trip test simulation

Diseño e implementación de algoritmos de unificación y matching para términos comprimidos.

OpenAIRE

Ramos Garrido, Lander

2011-01-01

Las gramáticas incontextuales se pueden utilizar como herramienta para comprimir palabras o términos. Estas gramáticas pueden permitir en casos extremos comprimir hasta un factor exponencial. Sobre este tipo de compresiones existen diferentes algoritmos que se pueden implementar sin necesidad de descomprimir la palabra o término dados. En este PFC diseñamos nuevos algoritmos probabilísticos que se puedan aplicar a este tipo de compresiones. Estos algoritmos mejoran el coste de los algoritmos ...

Algoritmo de reconocimiento de patrones basado en codificación fisiológica en cerebro de primates.

OpenAIRE

CASTEL BAIXAULI, ALEJANDRO

2017-01-01

El cerebro humano y animal es capaz de reconocer una gran cantidad de patrones, como por ejemplo caras, utilizando un limitado número de neuronas y algoritmos de procesamiento [Chang et al 2017]. Sin embargo, los algoritmos de reconocimiento de patrones utilizados en la actualidad, incluso aquellos basados en redes neuronales, requieren un alto número de operaciones y capacidad de computo. El objetivo del presente trabajo final de grado es adaptar, desarrollar y validar un algoritmo de re...

Best-estimate analysis development for BWR systems

International Nuclear Information System (INIS)

Sutherland, W.A.; Alamgir, M.; Kalra, S.P.; Beckner, W.D.

1986-01-01

The Full Integral Simulation Test (FIST) Program is a three pronged approach to the development of best-estimate analysis capability for BWR systems. An experimental program in the FIST BWR system simulator facility extends the LOCA data base and adds operational transients data. An analytical method development program with the BWR-TRAC computer program extends the modeling of BWR specific components and major interfacing systems, and improves numerical techniques to reduce computer running time. A method qualification program tests TRAC-B against experiments run in the FIST facility and extends the results to reactor system applications. With the completion and integration of these three activities, the objective of a best-estimate analysis capability has been achieved. (author)

Panorama of the BWR reactors - Evolution of the concept

Energy Technology Data Exchange (ETDEWEB)

Novotny, C.; Uhrig, E. [AREVA NP GmbH, Safety Engineering Department - PEPS-G (Germany)

2012-01-15

Nowadays, a fleet of more than 50 boiling water reactors (BWR) are in operation in the world. This article gives a short overview on the developments of nuclear power plants of the BWR type, with a focus on the European builds. It describes the technical bases from the early designs in the fifties, sketches the innovations of the sixties and seventies in the types BWR 69 and 72 (Baulinie 69 and 72) and gives an outlook of a possible next generation BWR. A promising approach in recent BWR developments is the the combination of passive safety systems with established design basis

BWR AXIAL PROFILE

International Nuclear Information System (INIS)

Huffer, J.

2004-01-01

The purpose of this calculation is to develop axial profiles for estimating the axial variation in burnup of a boiling water reactor (BWR) assembly spent nuclear fuel (SNF) given the average burnup of an assembly. A discharged fuel assembly typically exhibits higher burnup in the center and lower burnup at the ends of the assembly. Criticality safety analyses taking credit for SNF burnup must account for axially varying burnup relative to calculations based on uniformly distributed assembly average burnup due to the under-burned tips. Thus, accounting for axially varying burnup in criticality analyses is also referred to as accounting for the ''end effect'' reactivity. The magnitude of the reactivity change due to ''end effect'' is dependent on the initial assembly enrichment, the assembly average burnup, and the particular axial profile characterizing the burnup distribution. The set of bounding axial profiles should incorporate multiple BWR core designs and provide statistical confidence (95 percent confidence that 95 percent of the population is bound by the profile) that end nodes are conservatively represented. The profiles should also conserve the overall burnup of the fuel assembly. More background on BWR axial profiles is provided in Attachment I

Pronóstico puntos críticos de la serie temporal "consumo de energía eléctrica del sector industrial en la ciudad de Medellín,", usando algoritmos genéticos

Directory of Open Access Journals (Sweden)

Héctor Tabares

2007-01-01

Full Text Available Los algoritmos genéticos (AG están inspirados en el principio darwiniano de la evolución de las especies y en la genética. Son algoritmos probabilísticos que ofrecen un mecanismo de búsqueda paralela y adaptativa, basado en el principio de supervivencia de los más aptos y en la reproducción.Este artículo presenta una introducción a los fundamentos de los AG. También enseña el simulador software AG_UdeA desarrollado con un propósito didáctico para la enseñanza de los AG. El principal aporte consiste en la aplicación de los AG para pronosticar los puntos críticos de consumo de energía eléctrica del sector industrial de la ciudad de Medellín para un período de 24 horas.

GPE-BWR and the containment venting and filtering issue

International Nuclear Information System (INIS)

Palomo, J.; Santiago, J. de

1988-01-01

The Spanish Boiling Water Reactor Owner's Group (GPE-BWR) is formed by three utilities, owning four units: Santa Maria de Garona (46 MWe, BWR3, Mark I containment), Cofrentes (975 MWe, BWR6, Mark III containment) and Valdecaballeros (2x975 MWe, BWR6, Mark III containment) - all of the reactors having been supplied by General Electric. One of the GPE-BWR's several committees is the Safety and Licensing Committee, which follows up the evolution of severe accident topics and particularly the containment venting and filtering issue. In September 1987, the Consejo de Seguridad Nuclear (CSN), the Spanish Regulatory Body, asked the GPE-BWR to define its position on the installation of a containment venting system. The GPE-BWR created a Working Group which presented a Report on Containment Venting to the CSN in January 1987 gathered from: the US Nuclear Regulatory Commission (NRC); some US utilities; and several European countries, especially France, Germany and Sweden. CSN's review of the containment venting Report and the Action Plan proposed by the GPE-BWR finished in April 1988. The conclusion of the Report and the proposed Action Plan take into account the US NRC's identified open items on severe accidents and the R and D programs scheduled to close these items

Algunas relaciones entre algoritmos y resolución de problemas

Directory of Open Access Journals (Sweden)

Jaime Fonseca González

2012-04-01

Full Text Available Como parte del estudio realizado en la tesis de maestría “Uso de algoritmos en la resolución de problemas sobre isometrías del plano. Un estudio de caso”, se han identificado diversas formas en que se relacionan los algoritmos con la resolución de problemas, no sólo como procesos mecánicos, sino como fuente de información en el planteamiento de estrategias al resolver un problema, de modo que surgen algunas maneras en que se da esta relación, ejemplificando cada una de ellas con problemas cuyo objeto matemático central son las isometrías del plano.

Algoritmo de criptografia RSA: análise entre a segurança e velocidade

Directory of Open Access Journals (Sweden)

Rafael Santos Andrade

2012-12-01

Full Text Available O objetivo deste trabalho é abordar a relação existente entre a busca pela segurança de dados e a velocidade de codificação e decodificação do algoritmo de criptografia RSA, que utiliza um par de números inteiros como ‘chave’. Considerando o tamanho da chave como requisito de segurança, devido à dificuldade computacional de fatorar números inteiros extensos, simulamos estes processos, com o algoritmo implementado na linguagem de programação C, utilizando chaves aleatórias de 1024, 2048 e 4096 bits. Desta forma, observamos o tempo de processamento em função do tamanho das chaves, confrontando segurança e desempenho.Palavras-chave: criptografia; algoritmo RSA; segurança; desempenho.

Algoritmo para el cálculo de la velocidad media óptima en una ruta (ASGA

Directory of Open Access Journals (Sweden)

V. Corcoba Magaña

2014-10-01

Full Text Available Resumen: En este trabajo se propone un algoritmo para obtener la velocidad media óptima para ahorrar combustible y mejorar la seguridad. El algoritmo propuesto se basa en los algoritmos genéticos. El algoritmo emplea información sobre el entorno, la carretera y el vehículo para obtener la velocidad media que minimice el consumo de combustible sin incrementar drásticamente la duración del trayecto. Además, el algoritmo propuesto mejora la seguridad ya que adecua la velocidad a las condiciones de la vía. La información sobre el entorno se obtiene de servicios web y la información sobre el vehículo se obtiene a través del puerto OBD2. El algoritmo es validado en situaciones reales con incidentes de tráfico y sin ellos. Por otra parte, se analiza el impacto de la velocidad media y los incidentes de tráfico en las aceleraciones y su influencia en el consumo de combustible. Abstract: This paper proposes an algorithm for obtaining the optimal average speed to save fuel and improve safety. The proposed algorithm is based on genetic algorithms. The algorithm uses information about the environment, the road and the vehicle for obtaining the optimal average speed which it minimizes fuel consumption without dramatically increasing the travel time. Moreover, the proposed algorithm improves safety adapting vehicle speed to road conditions. The environment information is obtained from web services and vehicle information is obtained through the OBD2 port. The algorithm is validated in situations with and without incidents. In addition, we analyze the impact of the average speed and acceleration incidents and their impact on fuel consumption. Palabras clave: Conducción eficiente, Sistemas de ayuda a la conducción, Algoritmos Genéticos, Android, Sistemas Inteligentes de Transporte., Keywords: Eco-driving, Advanced Driver Assistance Systems, Genetic Algorithms, Android, Intelligent Transport System.

Algoritmo de tipo búsqueda tabú para un problema de programación de horarios universitarios vespertinos

Directory of Open Access Journals (Sweden)

Cristian David Oliva San Martín

2013-07-01

Full Text Available En este estudio se presenta un modelo de programación no lineal en variables enteras para un problema de programación de horarios universitarios vespertinos y se propone un algoritmo para su solución. El problema consiste en programar asignaturas en un horizonte de planificación considerando profesores, aulas, alumnos, días y un conjunto de restricciones, buscando minimizar la penalidad de no satisfacer los requerimientos de profesores, el número de cambios de sala y el número de periodos libres entre cursos. Se propone un modelo matemático y una implementación de un algoritmo de tipo búsqueda tabú. Para evaluar la efectividad del algoritmo se utilizaron casos de prueba con datos reales del Instituto IPEGE (Chile, en los que el algoritmo es capaz de obtener soluciones factibles en un tiempo razonable. Los parámetros de dicho algoritmo fueron calibrados con los casos de prueba, para posteriormente evaluar su desempeño. Se muestra que este obtuvo mejores soluciones que el método manual.

Algoritmo para el cálculo de cargas de trabajo

Directory of Open Access Journals (Sweden)

Mauricio Becerra Fernández

2016-04-01

Full Text Available Este artículo presenta el desarrollo un algoritmo para el cálculo de cargas de trabajo, soportado en el análisis de estudios similares y la teoría del estudio de métodos y tiempos. El algoritmo se propone como un elemento de soporte para el análisis del nivel trabajo asignado a un cargo específico o a un grupo de trabajadores en un área determinada, ya sea en la industria de manufactura o de servicios, lo que permita nivelar el volumen de trabajo asignado mediante la definición de diversas alternativas que promuevan el aumento de la productividad, favorezcan la mejora en el clima y las condiciones laborales.

Algoritmos rápidos de detecção de colisão broad phase utilizando KD-trees

OpenAIRE

Rocha, Rafael de Sousa

2010-01-01

Neste trabalho, três novos algoritmos rápidos de detecção de colisão broad phase, os quais utilizam a estrutura de particionamento espacial conhecida como KD-Tree, foram pro- postos e implementados: KDTreeSpace, DynamicKDTreeSpace e StatelessKDTreeSpace. Estes algoritmos foram integrados à biblioteca Open Dynamics Engine (ODE), responsável pelo cálculo do movimento dos objetos dinâmicos, como possíveis alternativas aos algoritmos de broad phase disponíveis nesta biblioteca. Os algori...

BWR stability analysis

International Nuclear Information System (INIS)

Valtonen, K.

1990-01-01

The objective of this study has been to examine TVO-I oscillation incident, which occured in February 22.1987 and to find out safety implications of oscillations in ATWS incidents. Calculations have been performed with RAMONA-3B and TRAB codes. RAMONA-3B is a BWR transient analysis code with three-dimencional neutron kinetics and nonequilibrium, nonhomogeneous thermal hydraulics. TRAB code is a one-dimencional BWR transient code which uses methods similar to RAMONA-3B. The results have shown that both codes are capable of analyzing of the oscillation incidents. Both out-of-phase and in-phase oscillations are possible. If the reactor scram fails (ATWS) during oscillations the severe fuel failures are always possible and the reactor core may exceed the prompt criticality

Optimización en la solución del problema inverso en geofísica usando el algoritmo de entrenamiento supervisado de Levenberg-Marquardt

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Figueredo Baez Yaqueline

2002-08-01

Full Text Available

In this work it is introduced a methodology for the supervised training of the Neural Networks, using Levenberg Marquardt's algorithm. This method is applied in gravimetry for the optimization in the convergence at the inverse problem.

En este trabajo se presenta una metodología para el entrenamiento supervisado de redes neuronales, utilizando el algoritmo de Levenberg-Marquardt. Este método se aplica en gravimetría para optimizar la convergencia en el problema de inversión.

Fundamentals of boiling water reactor (BWR)

International Nuclear Information System (INIS)

Bozzola, S.

1982-01-01

These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

BWR Refill-Reflood Program, Task 4.7 - model development: TRAC-BWR component models

International Nuclear Information System (INIS)

Cheung, Y.K.; Parameswaran, V.; Shaug, J.C.

1983-09-01

TRAC (Transient Reactor Analysis Code) is a computer code for best-estimate analysis for the thermal hydraulic conditions in a reactor system. The development and assessment of the BWR component models developed under the Refill/Reflood Program that are necessary to structure a BWR-version of TRAC are described in this report. These component models are the jet pump, steam separator, steam dryer, two-phase level tracking model, and upper-plenum mixing model. These models have been implemented into TRAC-B02. Also a single-channel option has been developed for individual fuel-channel analysis following a system-response calculation

Algoritmos de aproximação para problemas de empacotamento em faixa com restrições de descarregamento

OpenAIRE

Jefferson Luiz Moisés da Silveira

2011-01-01

Resumo: Neste trabalho estudamos problemas de empacotamento com restrições de descarregamento considerados NP-difíceis. Estes problemas possuem aplicações nas áreas de logística e roteamento. Assumindo a hipótese de que P ? NP, sabemos que não existem algoritmos eficientes para resolver tais problemas. Uma das abordagens consideradas para tratar tais problemas é a de algoritmos de aproximação, que são algoritmos eficientes (complexidade de tempo polinomial) e que geram soluções com garantia d...

Boiling water system of nuclear power plants (BWR)

International Nuclear Information System (INIS)

Martias Nurdin

1975-01-01

About 85% of the world electric generators are light water reactors. It shows that LWR is technologically and economically competitive with other generators. The Boiling Water Reactor (BWR) is one of the two systems in the LWR group. The techniques of BWR operation in several countries, especially low and moderate power BWR, are presented. The discussion is made in relation with the interconnection problems of electric installation in developing countries, including Indonesia, where the total electric energy installation is low. The high reliability and great flexibility of the operation of a boiling water reactor for a sufficiently long period are also presented. Component standardization for BWR system is discussed to get a better technological and economical performance for further development. (author)

CB-SMoT+: UNA EXTENSIÓN AL ALGORITMO CB-SMoT

Directory of Open Access Journals (Sweden)

FRANCISCO MORENO

2012-12-01

Full Text Available Una trayectoria es un registro de la evoluciónde la posición de un objeto móvil. Por ejemplo,un vehículo que se mueve en el espacio durante unintervalo de tiempo. Una trayectoria se representamediante una secuencia de observaciones que indicanla posición y el tiempo en el que fue tomada cadaobservación. CB-SMoT es un algoritmo que identificalas partes de una trayectoria durante las cuales elobjeto mantuvo una velocidad promedio por debajode un límite dado. En este artículo se propone unaextensión para dicho algoritmo que permite identificarlas partes de una trayectoria durante las cualesel objeto mantuvo una velocidad promedio entre observacionespor debajo de un límite dado. Esto posibilitala identificación, por ejemplo, de violacionesa un límite de velocidad que no son advertidas porel algoritmo original. Para el estudio se usó el sistemade gestión de bases de datos PostgreSQL y losalgoritmos se implementaron en su lenguaje de programación,llamado PL/pgSQL. Además, se hicieronexperimentos con 100 trayectorias de vehículos conel propósito de mostrar la utilidad y la viabilidad de lapropuesta.

Algoritmos evolutivos aplicados al diseño inteligente de parques eólicos

OpenAIRE

Bilbao, Martín; Alba, Enrique

2011-01-01

En este documento se muestran dos tipos de algoritmos metaheurísticos aplicados al diseño inteligente de un parque eólico, la idea básica es utilizar CHC y algoritmos genéticos para obtener una buena configuración de molinos dentro del parque de tal manera que maximicen la energía total producida y minimicen la cantidad de molinos utilizados. En este trabajo analizaremos dos casos de estudios con una distribución real de vientos de la patagonia argentina, un terreno irregular con restriccione...

Algoritmos de compresión paralela

OpenAIRE

Anderson, Alfredo; Dirazar, Delio

1997-01-01

El objetivo planteado inicialmente fue analizar la viabilidad de distribuir un compresor de datos en una red de procesadores. Además de elegir el algoritmo a implementar y definir alternativas de distribución debíamos seleccionar un lenguaje y un sistema operativo que soporten las herramientas de multiprocesamiento necesarias para la implementación de las versiones distribuidas. Nuestro primer paso fué realizar la implementación de dos compresores de datos basados en el mismo algortimo,...

Synergistic failure of BWR internals

International Nuclear Information System (INIS)

Ware, A. G.; Chang, T.Y.

1999-01-01

Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components

Utility experience with BWR-PSMS

International Nuclear Information System (INIS)

Bond, G.R.

1986-01-01

The BWR Power Shape Monitoring System (BWR-PSMS) has proven to be an effective and versatile tool for core monitoring. GPU Nuclear Corporation's (GPUN) Oyster Creek plant has been involved in the PSMS development since its inception, having been selected by EPRI as the initial demonstration site. Beginning with Cycle 10, Oyster Creek has been applying the BWR-PSMS as the primary core monitoring tool. Although the system has been in operation at Oyster Creek for the past several cycles, this is the first time the PSMS was used to monitor compliance to the plant technical specifications, to guide adherence to vendore fuel maneuvering recommendations and to develop data for certain performance records such as fuel burnup, isotopic accounting, etc. This paper will discuss the bases for the decision to apply PSMS as the fundamental core monitoring system, the experience in implementing the PSMS in this mode, activities currently underway or planned related to PSMS, and potential future extensions and applications of PSMS at Oyster Creek

EPRI BWR Water Chemistry Guidelines Revision

International Nuclear Information System (INIS)

Garcia, Susan E.; Giannelli, Joseph F.

2014-01-01

BWRVIP-190: BWR Water Chemistry Guidelines – 2008 Revision has been revised. The revision committee consisted of U.S. and non-U.S. utilities (members of the BWR Vessel and Internals Protection (BWRVIP) Mitigation Committee), reactor system manufacturers, fuel suppliers, and EPRI and industry experts. The revised document, BWRVIP-190 Revision 1, was completely reformatted into two volumes, with a simplified presentation of water chemistry control, diagnostic and good practice parameters in Volume 1 and the technical bases in Volume 2, to facilitate use. The revision was developed in parallel and in coordination with preparation of the Fuel Reliability Guidelines Revision 1: BWR Fuel Cladding Crud and Corrosion. Guidance is included for plants operating under normal water chemistry (NWC), moderate hydrogen water chemistry (HWC-M), and noble metal application (GE-Hitachi NobleChem™) plus hydrogen injection. Volume 1 includes significant changes to BWR feedwater and reactor water chemistry control parameters to provide increased assurance of intergranular stress corrosion cracking (IGSCC) mitigation of reactor materials and fuel reliability during all plant conditions, including cold shutdown (≤200°F (93°C)), startup/hot standby (>200°F (93°C) and ≤ 10%) and power operation (>10% power). Action Level values for chloride and sulfate have been tightened to minimize environmentally assisted cracking (EAC) of all wetted surfaces, including those not protected by hydrogen injection, with or without noble metals. Chemistry control guidance has been enhanced to minimize shutdown radiation fields by clarifying targets for depleted zinc oxide (DZO) injection while meeting requirements for fuel reliability. Improved tabular presentations of parameter values explicitly indicate levels at which actions are to be taken and required sampling frequencies. Volume 2 provides the technical bases for BWR water chemistry control for control of EAC, flow accelerated corrosion

Algoritmo genético e espectroscopia no infravermelho - algumas aplicações na indústria cosmética

OpenAIRE

Marcos Coelho Amendola

2007-01-01

Este trabalho discute o desenvolvimento de um algoritmo genético escrito em linguagem VBA para Excel e suas aplicações. O algoritmo elaborado foi utilizado em combinação com a técnica de FTIR-ATR para o desenvolvimento de metodologias aplicáveis na indústria cosmética e de saneantes, tais como a quantificação de surfactantes e bactericidas. Algumas modificações introduzidas no algoritmo foram estudadas através das aplicações selecionadas, destacando-se a introdução de técnicas de paralelismo ...

Análise de desempenho de algoritmos evolutivos no domínio do futebol de robôs

OpenAIRE

Eduardo Sacogne Fraccaroli

2010-01-01

Muitos problemas de otimização em ambientes multiagentes utilizam os algoritmos evolutivos para encontrar as melhores soluções. Uma das abordagens mais utilizadas consiste na aplicação de um algoritmo genético, como alternativa aos métodos tradicionais, para definir as ações dos jogadores em um time de futebol de robôs. Entretanto, conforme relatado na literatura, há inúmeras possibilidades e formas de se aplicar um algoritmo genético no domínio do futebol de robôs. Assim sendo, neste trabalh...

BWR Services maintenance training program

International Nuclear Information System (INIS)

Cox, J.H.; Chittenden, W.F.

1979-01-01

BWR Services has implemented a five-phase program to increase plant availability and capacity factor in operating BWR's. One phase of this program is establishing a maintenance training program on NSSS equipment; the scope encompasses maintenance on both mechanical equipment and electrical control and instrumentation equipment. The program utilizes actual product line equipment for practical Hands-on training. A total of 23 formal courses will be in place by the end of 1979. The General Electric Company is making a multimillion dollar investment in facilities to support this training. These facilities are described

Behaviour of the reactivity for BWR fuel cells; Comportamiento de la reactividad para celdas de combustible BWR

Energy Technology Data Exchange (ETDEWEB)

Gonzalez, J. A.; Alonso, G.; Delfin, A.; Vargas, S. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: galonso@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico)

2011-11-15

In this work the behaviour of the reactivity of a fuel assembly type BWR was studied, the objective is to obtain some expressions that consider the average enrichment of U-235 and the gadolinium concentration like a function of the fuel cells burnt. Also, the applicability of the lineal reactivity model was analyzed for fuel cells type BWR. The analysis was carried out with the CASMO-4 code. (Author)

Moderator temperature coefficient in BWR core

International Nuclear Information System (INIS)

Naito, Yoshitaka

1977-01-01

Temperature dependences of infinite multiplication factor k sub(infinity) and neutron leakage from the core must be examined for estimation of moderator temperature coefficient. Temperature dependence on k sub(infinity) has been investigated by many researchers, however, the dependence on neutron leakage of a BWR with cruciformed control rods has hardly been done. Because there are difficulties and necessity on calculations of three space dimensional and multi-energy groups neutron distribution in a BWR core. In this study, moderator temperature coefficients of JPDR-II (BWR) core were obtained by calculation with DIFFUSION-ACE, which is newly developed three-dimensional multi-group computer code. The results were compared with experimental data measured from 20 to 275 0 C of the moderator temperature and the good agreement was obtained between calculation and measurement. In order to evaluate neutron leakage from the core, the other two calculations were carried out, adjusting criticality by uniform absorption rate and by material buckling. The former underestimated neutron leakage and the latter overestimated it. Discussion on the results shows that in order to estimate the temperature coefficient of BWR, neutron leakage must be evaluated precisely, therefore the calculation at actual pattern of control rods is necessary. (auth.)

BWR stability analysis at Brookhaven National Laboratory

International Nuclear Information System (INIS)

Wulff, W.; Cheng, H.S.; Mallen, A.N.; Rohatgi, U.S.

1991-01-01

Following the unexpected, but safely terminated, power and flow oscillations in the LaSalle-2 Boiling Water Reactor (BWR) on March 9, 1988, the Nuclear Regulatory Commission (NRC) Offices of Nuclear Reactor Regulation (NRR) and of Analysis and Evaluation of Operational Data (AEOD) requested that the Office of Nuclear Regulatory Research (RES) carry out BWR stability analyses, centered around fourteen specific questions. Ten of the fourteen questions address BWR stability issues in general and are dealt with in this paper. The other four questions address local, out-of-phase oscillations and matters of instrumentation; they fall outside the scope of the work reported here. It was the purpose of the work documented in this report to answer ten of the fourteen NRC-stipulated questions. Nine questions are answered by analyzing the LaSalle-2 instability and related BWR transients with the BNL Engineering Plant Analyzer (EPA) and by performing an uncertainty assessment of the EPA predictions. The tenth question is answered on the basis of first principles. The ten answers are summarized

IMPLEMENTACIÓN EN HARDWARE DE UN SVPWM EN UN SOFT-CORE NIOS II PARTE II: ALGORITMO DEL SVPWM

Directory of Open Access Journals (Sweden)

Jorge Rivera

2009-01-01

Full Text Available En este trabajo se presentan los detalles de la moduclación por ancho de pulso en el espacio vectorial (SVPWM por sus siglas en inglés. Primero se plantea el escenario en donde es más usado este algoritmo, es decir, en el control de motores eléctricos trifásicos. Se explica que el algoritmo del SVPWM sirve para manejar los inversores, además se indica en qué marco de referencia deben estar las acciones de control que el algortimo del SVPWM acepta, para finalmente presentar el algoritmo de una manera clara y concisa.

Algoritmos y gramáticas : un capitulo apasionante de la teoría cognitiva

OpenAIRE

Pérez Alcázar, Jesús Hernando

2012-01-01

DOS GRANDES PROBLEMASUno de los grandes pioneros de la teoría cognitiva es el profesor norteamericano Noam Chomsky. En este ensayo presentaremos algunos de sus muy numerosos aportes al desarrollo de esta nueva teoría, limitando la atención a las relaciones entre gramaticas y algoritmos. Una manera muy sugestiva de abordar las relaciones entre los algoritmos matemáticos y las gramáticas de la teoría lingüística, la ofrece el propio Noam Chomsky en su importante trabajo "El conocimiento del len...

Paralelización del Algoritmo Expectación–Maximización Utilizando OpenCL

Directory of Open Access Journals (Sweden)

Ernesto Insua-Suárez

2014-12-01

Full Text Available Actualmente, las organizaciones y empresas almacenan grandes volúmenes de datos para lograr sus propósitos. Una de las variantes para obtener información valiosa consiste en el empleo de la Minería de datos. Dentro de esta, existen diferentes tareas, una de ellas es el agrupamiento. En esta tarea los datos se agrupan según sus semejanzas entre si y diferencias con elementos de otros grupos. Dentro de los algoritmos que realizan estos agrupamientos se encuentra Expectación-Maximización, el cual presenta elevados tiempos de ejecución en la medida que aumenta el tamaño de los datos. En el presente artículo se discute acerca de la paralelización del algoritmo, utilizando técnicas de programación paralela. El diseño del algoritmo propuesto se basa en el uso de las tarjetas de procesamiento gráfico, GPU. OpenCL, lenguaje empleado para la programación en arquitecturas híbridas, permite aprovechar las arquitecturas de hardware disponibles, con lo que se logra disminuir el tiempo de ejecución de la implementación realizada. La razón principal por lo cual es posible mejorar este tiempo se debe a la cantidad de procesos paralelos que se pueden lanzar en hilos de procesamientos independientes. Para el logro de los resultados descritos se integran conocimientos del campo de la Minería de datos y la Computación Paralela y Distribuida. Como parte de esta investigación, se realizó una implementación del algoritmo utilizando las bibliotecas de OpenCL, para disminuir su tiempo de ejecución. La implementación logra disminuir en un 82% la implementación secuencial. Esto significa que el algoritmo paralelo se ejecuta 5,5 veces más rápido que su correspondiente implementación secuencial.

Residual stress analysis in BWR pressure vessel attachments

International Nuclear Information System (INIS)

Dexter, R.J.; Leung, C.P.; Pont, D.

1992-06-01

Residual stresses from welding processes can be the primary driving force for stress corrosion cracking (SCC) in BWR components. Thus, a better understanding of the causes and nature of these residual stresses can help assess and remedy SCC. Numerical welding simulation software, such as SYSWELD, and material property data have been used to quantify residual stresses for application to SCC assessments in BWR components. Furthermore, parametric studies using SYSWELD have revealed which variables significantly affect predicted residual stress. Overall, numerical modeling techniques can be used to evaluate residual stress for SCC assessments of BWR components and to identify and plan future SCC research

BWR Refill-Reflood Program. Final report

International Nuclear Information System (INIS)

Myers, L.L.

1983-09-01

The BWR Refill-Reflood Program is part of the continuing Loss of Coolant Accident (LOCA) research in the United States which is jointly sponsored by the Nuclear Regulatory Commission, the Electric Power Research Institute, and the General Electric Company. The current program expanded the focus of this research to include full scale experimental evaluations of multidimensional and multichannel effects during system refill. The program has also made major contributions to the BWR version of the Transient Reactor Analysis Code (TRAC) which has been developed cooperatively with the Idaho National Engineering Laboratory (INEL) for application to BWR transients. A summary description of the complete program is provided including the principal findings and main conclusions of the program. The results of the program have shown that multidimensional and parallel channel effects have the potential to significantly improve the system response over that observed in single channel tests

Algoritmos de previsión a corto plazo en base a radar meteorológico

OpenAIRE

Campeny Mora, Mercè

2008-01-01

Las consecuencias de las inundaciones suelen ser devastadoras. Los sistemas de protección tradicionales se complementan cada vez más con sistemas de prevención y alerta de inundaciones. En este sentido, el radar meteorológico y los algoritmos de previsión adquieren una importancia clave. El objetivo de la tesina es la mejora de los algoritmos de previsión en base a radar meteorológico a partir del estudio de distintos factores que intervienen.

BWR Steam Dryer Alternating Stress Assessment Procedures

Energy Technology Data Exchange (ETDEWEB)

Morante, R. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hambric, S. A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ziada, S. [Brookhaven National Lab. (BNL), Upton, NY (United States)

2016-12-01

This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

Modelos de pérdida de masa de acero por corrosión atmosférica en Colombia usando inteligencia computacional

Directory of Open Access Journals (Sweden)

Esteban Velilla

2009-01-01

Full Text Available Con el fin de clasificar la corrosividad de las diferentes atmósferas colombianas, como parte de un proyecto de investigación extenso [1], se expusieron placas de acero al carbono en 21 estaciones distribuidas a lo largo de la infraestructura eléctrica del país (líneas de transmisión y subestaciones. En estas estaciones se midieron entre otros, el tiempo de humectación y la deposición de sulfatos y cloruros durante 12 meses; además, bimensualmente se tomaban placas de acero para medir en laboratorio la pérdida de masa sufrida por estas durante el tiempo de exposición. La clasificación de las 21 estaciones se hizo en 4 grupos, considerando: el tiempo de humectación, contenidos de cloruros y sulfatos, la altura sobre el nivel del mar y el tiempo de exposición de las placas; variables consideradas linealmente independientes según la técnica de descomposición en valores singulares (SVD realizada. El criterio utilizado para la clasificación fue el de similitud de las variables utilizando la norma Euclidiana considerada en la red neuronal no supervisada tipo Kohonen. Adicionalmente, se implementaron modelos para la pérdida de masa del acero para cada uno de los grupos usando redes neuronales (RN tipo Feed-Forward, definiéndose como entradas las variables antes mencionadas y como única salida la pérdida de masa. Complementariamente se presenta una comparación entre el modelo de RN para el grupo 1, con otros modelos obtenidos usando Algoritmos Genéticos (AG y el método Simplex.

Simplified distributed parameters BWR dynamic model for transient and stability analysis

International Nuclear Information System (INIS)

Espinosa-Paredes, Gilberto; Nunez-Carrera, Alejandro; Vazquez-Rodriguez, Alejandro

2006-01-01

This paper describes a simplified model to perform transient and linear stability analysis for a typical boiling water reactor (BWR). The simplified transient model was based in lumped and distributed parameters approximations, which includes vessel dome and the downcomer, recirculation loops, neutron process, fuel pin temperature distribution, lower and upper plenums reactor core and pressure and level controls. The stability was determined by studying the linearized versions of the equations representing the BWR system in the frequency domain. Numerical examples are used to illustrate the wide application of the simplified BWR model. We concluded that this simplified model describes properly the dynamic of a BWR and can be used for safety analysis or as a first approach in the design of an advanced BWR

Software para la enseñanza-aprendizaje de algoritmos estructurados

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Arellano Pimentel, J. Jesús

2012-12-01

Full Text Available Actualmente existen diversas herramientas de software que sirven como recurso didáctico en la enseñanza-aprendizaje de algoritmos estructurados a nivel superior. Sin embargo, la gran mayoría solo hace énfasis en el diseño y prueba de los algoritmos. En el presente trabajo se propone un nuevo software, con base en la heurística de resolución de problemas de Polya, que da soporte a las fases de análisis y planteamiento del problema, además del diseño y traza completa de la prueba. Se realiza un análisis cualitativo entre el software propuesto y otras tres herramientas ampliamente utilizadas a nivel superior. La estructura y funcionalidad del software propuesto contribuye a que el estudiante adquiera, practique y ejercite la capacidad de resolver problemas de forma metódica a través de soluciones algorítmicas estructuradas.

ABB advanced BWR and PWR fuel

International Nuclear Information System (INIS)

Junkrans, S.; Helmersson, S.; Andersson, S.

1999-01-01

Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

BWR control blade replacement strategies

Energy Technology Data Exchange (ETDEWEB)

Kennard, M W [Stoller Nuclear Fuel, NAC International, Pleasantville, NY (United States); Harbottle, J E [Stoller Nuclear Fuel, NAC International, Thornbury, Bristol (United Kingdom)

2000-02-01

The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B{sub 4}C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

BWR control blade replacement strategies

International Nuclear Information System (INIS)

Kennard, M.W.; Harbottle, J.E.

2000-01-01

The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B 4 C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

Burnup credit feasibility for BWR spent fuel shipments

International Nuclear Information System (INIS)

Broadhead, B.L.

1990-01-01

Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent of fuel casks used for transportation and storage. Analyses 1 have shown the feasibility estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This paper summarizes the extension of the previous PWR feasibility assessments to boiling water reactor (BWR) fuel. As with the PWR analysis, the purpose was not verification of burnup credit (see ref. 2 for ongoing work in this area) but a reasonable assessment of the feasibility and potential gains from its use in BWR applications. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. The method includes characterization of a typical pin-cell spectrum, using a one-dimensional (1-D) model of a BWR assembly. The calculated spectrum allows burnup-dependent few-group material constants to be generated. Point depletion methods were then used to obtain the time-varying characteristics of the fuel. These simple methods were validated, where practical, with multidimensional methods. 6 refs., 1 tab

Safety analysis of thorium-based fuels in the General Electric Standard BWR

International Nuclear Information System (INIS)

Colby, M.J.; Townsend, D.B.; Kunz, C.L.

1980-06-01

A denatured (U-233/Th)O 2 fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO 2 fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O 2 -fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O 2 -fueled BWR should perform similar to a UO 2 -fueled BWR under all operating conditions. A (Pu/Th)O 2 -fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO 2 -fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths

Application of the MOVE algorithm for the identification of reduced order models of a core of a BWR type reactor; Aplicacion del algoritmo MOVE para la identificacion de modelos de orden reducido del nucleo de un reactor tipo BWR

Energy Technology Data Exchange (ETDEWEB)

Victoria R, M.A.; Morales S, J.B. [UNAM, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: angelvr@gmail.com

2005-07-01

Presently work is applied the modified algorithm of the ellipsoid of optimal volume (MOVE) to a reduced order model of 5 differential equations of the core of a boiling water reactor (BWR) with the purpose of estimating the parameters that model the dynamics. The viability is analyzed of carrying out an analysis that calculates the global dynamic parameters that determine the stability of the system and the uncertainty of the estimate. The modified algorithm of the ellipsoid of optimal volume (MOVE), is a method applied to the parametric identification of systems, in particular to the estimate of groups of parameters (PSE for their initials in English). It is looked for to obtain the ellipsoid of smaller volume that guarantees to contain the real value of the parameters of the model. The PSE MOVE is a recursive identification method that can manage the sign of noise and to ponder it, the ellipsoid represents an advantage due to its easy mathematical handling in the computer, the results that surrender are very useful for the design of Robust Control since to smaller volume of the ellipsoid, better is in general the performance of the system to control. The comparison with other methods presented in the literature to estimate the reason of decline (DR) of a BWR is presented. (Author)

BWR vessel and internals project (BWRVIP)

International Nuclear Information System (INIS)

Bilanin, W.J.; Dyle, R.L.

1996-01-01

Recent Boiling Water Reactor (BWR) inspections indicate that Intergranular Stress Corrosion Cracking (IGSCC) is a significant technical issue for some BWR internals. IN response, the Boiling Water Reactor Vessel and Internals Project (BWRVIP) was formed by an associated of domestic and international utilities which own and operate BWRs. The project is identifying or developing generic, cost-effective strategies for managing degradation of reactor internals from which each utility can select the alternative most appropriate for their plant. The Electric Power Research Institute manages the technical program, implementing the utility defined programs. The BWRVIP is organized into four technical tasks: Assessment, Inspection, Repair and Mitigation. An Integration task coordinates the work. The goal of the Assessment task is to develop methodologies for evaluation of vessel and internal components in support of decisions for operation, inspection, mitigation or repair. The goal of the Inspection task is to develop and assess effective and predictable inspection techniques which can be used to determine the condition of BWR vessel and internals that are potentially susceptible to service-related SCC degradation. The goal of the Repair task is to assure the availability of cost-effective repair/replacement alternatives. The goal of the Mitigation task is to develop and demonstrate countermeasures for SCC degradation. This paper summarizes the BWRVIP approach for addressing BWR internals SCC degradation and illustrates how utilities are utilizing BWRVIP products to successfully manage the effect of SCC on core shrouds

Strategies of operation cycles in BWR type reactors

International Nuclear Information System (INIS)

Molina, D.; Sendino, F.

1996-01-01

The article analyzes the operation cycles in BWR type reactors. The cycle size of operation is the consequence on the optimization process of the costs with the technical characteristics of nuclear fuel and the characteristics of demand and production. The authors analyze the cases of Garona NP and Cofrentes NP, both with BWR reactors. (Author)

Review of international solutions to NEACRP benchmark BWR lattice cell problems

International Nuclear Information System (INIS)

Halsall, M.J.

1977-12-01

This paper summarises international solutions to a set of BWR benchmark problems. The problems, posed as an activity sponsored by the Nuclear Energy Agency Committee on Reactor Physics, were as follows: 9-pin supercell with central burnable poison pin, mini-BWR with 4 pin-cells and water gaps and control rod cruciform, full 7 x 7 pin BWR lattice cell with differential U 235 enrichment, and full 8 x 8 pin BWR lattice cell with water-hole, Pu-loading, burnable poison, and homogenised cruciform control rod. Solutions have been contributed by Denmark, Japan, Sweden, Switzerland and the UK. (author)

Improvement for BWR operator training, 3

International Nuclear Information System (INIS)

Noji, Kunio; Toeda, Susumu; Saito, Genhachi; Suzuki, Koichi

1990-01-01

BWR Operator Training Center Corporation (BTC) is conducting training for BWR plant operators using Full-scope Simulators. There are several courses for individual operators and one training course for shift crew (Family Training Course) in BTC. Family Training is carried out by all members of the operating shift-crew. BTC has made efforts to improve the Family Training in order to acquire more effective training results and contribute to up-grade team performance of all crews. This paper describes some items of our efforts towards Family Training improvement. (author)

Assessment of two BWR accident management strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Petek, M.

1991-01-01

Candidate mitigative strategies for management of in-vessel events during the late phase (after core degradation has occurred) of postulated BWR severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for additional assessment. The first is a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertains to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose is to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies have been performed during 1991 under the auspices of the Detailed Assessment of BWR In-Vessel Strategies Program. This paper provides a discussion of the motivation for and purpose of these strategies and the potential for their success. 33 refs., 9 figs

Kinematics of two-phase mixture level motion in BWR pressure vessels

International Nuclear Information System (INIS)

Wulff, W.; Cheng, H.S.; Mallen, A.N.; Stritar, A.

1985-01-01

A model is presented for predicting two-phase mixture level elevations in BWR systems. The model accounts for the particular geometry and conditions in a BWR system during Small-Break Loss of Coolant Accidents. The model presented here is particularly suitable for efficient, high-speed simulations on small minicomputers. The model has been implemented and tested. Results are shown from BWR ATWS simulations

Improvement for BWR operator training

International Nuclear Information System (INIS)

Tsuchiya, Toshio; Masuda, Hisao; Isono, Tomoyuki; Noji, Kunio; Togo, Toshiki

1989-01-01

BWR Operator Training Center Corporation (BTC) was established in April 1971 for the purpose of training the operators from all BWR utilities in Japan. Since April 1974, more than 2600 operators and 1000 shift teams have been trained with the full-scope simulators in BTC up to the end of March 1988. To get the satisfactory results of the training, BTC has been making every effort to improve the facilities, the training materials, the instruction methods and the curricula. In this paper, such a series of recent improvements in the instruction methods and the curricula are presented that are effective to expand the knowledge and to improve the skills of middle or senior class operators. (author)

Diseño óptimo de un sistema de distribución de agua (SDA aplicando el algoritmo Simulated Annealing (SA

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Maikel Méndez-Morales

2014-09-01

Full Text Available En este artículo se presenta la aplicación del algoritmo Simulated Annealing (SA en el diseño óptimo de un sistema de distribución de agua (SDA. El SA es un algoritmo metaheurístico de búsqueda, basado en una analogía entre el proceso de recocido en metales (proceso controlado de enfriamiento de un cuerpo y la solución de problemas de optimización combinatorios. El algoritmo SA, junto con diversos modelos matemáticos, ha sido utilizado exitosamente en el óptimo diseño de SDA. Como caso de estudio se utilizó el SDA a escala real de la comunidad de Marsella, en San Carlos, Costa Rica. El algoritmo SA fue implementado mediante el conocido modelo EPANET, a través de la extensión WaterNetGen. Se compararon tres diferentes variaciones automatizadas del algoritmo SA con el diseño manual del SDA Marsella llevado a cabo a prueba y error, utilizando únicamente costos unitarios de tuberías. Los resultados muestran que los tres esquemas automatizados del SA arrojaron costos unitarios por debajo del 0.49 como fracción, respecto al costo original del esquema de diseño ejecutado a prueba y error. Esto demuestra que el algoritmo SA es capaz de optimizar problemas combinatorios ligados al diseño de mínimo costo de los sistemas de distribución de agua a escala real.

Application of gadolinia credit to cask transportation of BWR-STEP3 SFAs

International Nuclear Information System (INIS)

Kikuchi, Tsukasa; Mitsuhashi, Ishi; Ito, Dai-ichiro; Nakamura, Yu

2003-01-01

Instead of the fresh-fuel assumption, the application of gadolinia credit to cask transportation of BWR SFAs is studied. Its efficacy for BWR-STEP2 SFAs had already been estimated. This paper reports on the application of gadolinia credit to cask transportation of BWR-STEP3 SFAs. (author)

Comprensión de los algoritmos de las operaciones aritméticas en estudiantes de Magisterio

OpenAIRE

Salinas, María Jesús

2003-01-01

Partiendo de la creencia de que las matemáticas escolares deben formar parte del componente del conocimiento sobre las matemáticas del futuro maestro, se hizo un estudio sobre dos campos sistema de numeración decimal y algoritmos usuales de las operaciones aritméticas. En esta comunicación nos centramos solamente en el campo correspondiente a los algoritmos. Presentaremos los resultados generales y un análisis de las respuestas de 82 protocolos de una muestra de 467 sujetos correspondiente al...

DISEÑO DE UN MICROSISTEMA PROGRAMABLE PARA EFECTOS DE AUDIO DIGITAL USANDO FPGAS

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John Michael Espinosa Durán

Full Text Available Este artículo describe el diseño de un microsistema programable para el procesamiento de efectos de audio digital implementado en un FPGA. El microsistema es diseñado usando un procesador de propósito específico y reconfigurable, un banco de RAMs y una interfaz gráfica de usuario basada en una pantalla táctil LCD. El procesador es diseñado usando 15 efectos de audio basados en retardos y procesamiento en el dominio dinámico y de la frecuencia. Los efectos son diseñados usando Megafunciones y el compilador FIR de Quartus II, son simulados en Simulink5 usando DSP Builder6, y son configurados utilizando una interfaz gráfica de usuario. El microsistema programable es implementado en el sistema de desarrollo DE2-70, y su funcionamiento es verificado usando un reproductor MP3 y un parlante. Adicionalmente, el microsistema permite la generación de efectos con alta fidelidad usando una tasa de muestreo máxima de 195.62 MSPS, y puede ser embebido en un SoC.

Gerando orientações acíclicas com algoritmos probabilísticos distribuídos

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Gladstone M. Arantes Jr

2005-12-01

Full Text Available Este artigo apresenta um novo algoritmo distribuído probabilístico para a geração de orientações acíclicas em um sistema distribuído anônimo de topologia arbitrária. O algoritmo é analisado tanto em termos de correção e complexidade esperada quanto velocidade de convergência. Em particular, é demonstrado que este novo algoritmo, chamado Alg-Arestas, é capaz de produzir, com alta probabilidade, orientações acíclicas quase instantaneamente, isto é, em menos de dois passos. Duas aplicações para essa forma de quebra de simetria serão discutidas: (i inicialização do Escalonamento por Reversão de Arestas (ERA, um simples e poderoso algoritmo de escalonamento distribuído, e (ii uma estratégia de distribuição de uploads em redes de computadores.This paper presents a new randomized distributed algorithm for the generation of acyclic orientations upon anonymous distributed systems of arbitrary topology. This algorithm is analyzed in terms of correctness and complexity as well as its convergence rate. In particular, it is shown that this new algorithm, called Alg-Arestas, is able to produce, with high probability, acyclic orientations quasi instantaneously, i.e., in less than two steps. Two applications of this form of symmetry breaking will be discussed: (i initialization of Scheduling by Edge Reversal (SER, a simple and powerful distributed scheduling algorithm, and (ii a strategy for distributed uploading in computer networks.

ALGORITMO PARA LA SOLUCIÓN NUMÉRICA DE SISTEMAS DE ECUACIONES NO LINEALES MEDIANTE UNA ESTRATEGIA DE OPTIMIZACIÓN GLOBAL BASADA EN ANÁLISIS DE INTERVALOS ALGORÍTMO PARA A SOLUÇÃO NUMÉRICA DE SISTEMAS DE EQUAÇÕES NÃO LINEARES MEDIANTE UMA ESTRATÉGIA DE OTIMIZAÇÃO GLOBAL BASEADA EM ANÁLISE DE INTERVALOS AN ALGORITHM FOR NUMERICAL SOLUTION OF NONLINEAR EQUATIONS SYSTEMS USING A STRATEGY OF GLOBAL OPTIMIZATION BASED ON INTERVAL ANALYSIS

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Luis Antonio Gómez

2012-12-01

Full Text Available En este artículo se presenta un algoritmo para la solución numérica de sistemas de ecuaciones no lineales. Para este propósito el sistema de ecuaciones se convierte en una función de valor real para luego ser minimizada, en el sentido global, en un dominio inicial dado (una caja en n usando análisis de intervalos. El algoritmo diseñado tiene la capacidad de determinar la existencia o no de soluciones al sistema de ecuaciones en una caja dada. Las soluciones del sistema de ecuaciones, si existen dentro de la caja dada, son expresadas mediante encerramientos por subcajas cuyo tamaño es menor que la exactitud establecida. No hay restricción acerca de la relación entre el número de ecuaciones y el número de incógnitas del sistema. Se realiza además un análisis de la convergencia del algoritmo y se muestran los resultados de su aplicación para algunos problemas de prueba.Em este artigo apresenta-se um algoritmo para a solução numérica de sistemas de equações não lineares. Para este propósito o sistema de equações converte-se em uma função de valor real para logo ser minimizada, no sentido global, em um domínio inicial dado (uma caixa em n usando análise de intervalos. O algoritmo desenhado tem a capacidade de determinar a existência ou não de soluções ao sistema de equações em uma caixa dada. As soluções do sistema de equações, se existirem dentro da caixa dada, são expressas mediante fechamentos por subcaixas cujo tamanho é menor que a exatidão estabelecida. Não há restrição a respeito da relação entre o número de equações e o número de incógnitas do sistema. Realiza-se ademais uma análise da convergência do algoritmo e mostram-se os resultados de sua aplicação para alguns problemas de prova.In this paper an algorithm for the numerical solution of nonlinear equations systems is presented. For this purpose the system of equations becomes a function of real value which will be minimized, in the

VARIACION DE PARAMETROS DE CRIPTOGRAFIA CON CURVAS ELIPTICAS USADOS EN LA FIRMA DIGITAL DE DATOS SOBRE UNA RED DE SENSORES INALAMBRICOS

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Javier Omar Contreras Rodriguez

2016-09-01

Full Text Available En la actualidad, el auge de las aplicaciones de las redes de sensores inalámbricos (WSN = Wireless Sensors Networks está generando una gran cantidad de información de carácter sensible que requiere un manejo confiable mediante la implementación de sistemas de seguridad de los datos compatibles con la naturaleza de estas redes. En ese sentido, cada vez más aumenta el interés por el uso de algoritmos de criptografía de clave pública con curvas elípticas (ECC = Elliptic Curve Cryptography como una alternativa de menor consumo de recursos computacionales comparado con los algoritmos tradicionalmente usados, como por ejemplo, RSA (Rivest-Shamir-Adleman, Diffie-Hellman, otros. En este artículo, se revisa la construcción de un prototipo de sistema de seguridad usando ECC para la firma digital de datos (ECDSA = Elliptic Curve Digital Signature Algorithm usando un control lógico basado en redes definidas por software (SDN = Software Defined Networking para el control de funcionalidades básicas y que permita ajustar en tiempo real los parámetros del algoritmo ECDSA según el tipo de aplicación de la WSN.

Seismic risk assessment of a BWR: status report

International Nuclear Information System (INIS)

Chuang, T.Y.; Bernreuter, D.L.; Wells, J.E.; Johnson, J.J.

1985-02-01

The seismic risk methodology developed in the US NRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant, a pressurized water reactor (PWR). A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models was developed and analyzed. The SSMRP methodology can equally be applied to a boiling water reactor (BWR). To demonstrate its applicability, to identify fundamental differences in seismic risk between a PWR and a BWR, and to provide a basis of comparison of seismic risk between a PWR and a BWR when analyzed with comparable methodology and assumptions, a seismic risk analysis is being performed on the LaSalle County Station nuclear power plant

Machine vision system for weed detection using image filtering in vegetables crops

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Camilo Andrés Pulido-Rojas

2016-01-01

Full Text Available El presente trabajo expone un sistema de visión de máquina para la detección de maleza en cultivos de hortalizas, usando imágenes exteriores, evadiendo problemas de iluminación y nitidez durante la etapa de adquisición, ya que el presente desarrollo será un módulo para un robot móvil removedor de maleza con una cámara oscura para controlar las condiciones de luz. El enfoque de diseño se enmarca en el desarrollo de un algoritmo útil para discriminar maleza usando filtros en la imagen extrayendo características de color y área, luego se implementa un proceso de etiquetado para cada objeto en la escena. Finalmente, una clasificación basada en área es propuesta incluyendo el cálculo de los índices de sensibilidad, especificidad, valores predictivos positivos y negativos con el fin de evaluar el rendimiento del algoritmo.

BWR plant analyzer development at BNL

International Nuclear Information System (INIS)

Cheng, H.S.; Wulff, W.; Mallen, A.N.; Lekach, S.V.; Stritar, A.; Cerbone, R.J.

1985-01-01

Advanced technology for high-speed interactive nuclear power plant simulations is of great value for timely resolution of safety issues, for plant monitoring, and for computer-aided emergency responses to an accident. Presented is the methodology employed at BNL to develop a BWR plant analyzer capable of simulating severe plant transients at much faster than real-time process speeds. Five modeling principles are established and a criterion is given for selecting numerical procedures and efficient computers to achieve the very high simulation speeds. Typical results are shown to demonstrate the modeling fidelity of the BWR plant analyzer

Development of new irradiation facility for BWR safety research

International Nuclear Information System (INIS)

Okada, Yuji; Magome, Hirokatsu; Iida, Kazuhiro; Hanawa, Hiroshi; Ohmi, Masao

2013-01-01

In JAEA (Japan Atomic Energy Agency), about the irradiation embrittlement of the reactor pressure vessel and the stress corrosion cracking of reactor core composition apparatus concerning the long-term use of the light water reactor (BWR), in order to check the influence of the temperature, pressure, and water quality, etc on BWR condition. The water environmental control facility which performs irradiation assisted stress corrosion-cracking (IASCC) evaluation under BWR irradiation environment was fabricated in JMTR (Japan Materials Testing Reactor). This report is described the outline of manufacture of the water environmental control facility for doing an irradiation test using the saturation temperature capsule after JMTR re-operation. (author)

Análisis Comparativo de Algoritmos de Minería de Subgrafos Frecuentes

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Santiago Bianco

2016-06-01

Full Text Available Gracias a las posibilidades que ofrecen a la hora de representar información y la abstracción conceptual que permiten manejar, los grafos son ampliamente utilizados en investigaciones relacionadas con la informática. A medida que se fueron incrementando las aplicaciones de estas estructuras, la complejidad de los elementos a representar y el volumen de información manejado, aparece la necesidad de utilizar procesos eficientes para extraer información o patrones ocultos en esa gran masa de datos, por lo que se comienza a aplicar la minería de grafos. Dentro las técnicas de minería de grafos se encuentra la búsqueda de subgrafos frecuentes, utilizada para reconocer subestructuras comunes entre un conjunto de grafos. En los últimos años se han llevado a cabo varias investigaciones que resuelven este problema, generando algoritmos diversos aplicando distintos enfoques, entre los cuales se encuentran el FSG, FFSM, gSpan y GASTON. El objetivo de este artículo es analizar el comportamiento de estos algoritmos a través de distintos experimentos diseñados para identificar si existe un algoritmo superior al resto y, en caso de que no lo haya, poder definir en qué escenarios es más recomendable la elección de cada uno.

Interpretation of incore noise measurements in BWR's

International Nuclear Information System (INIS)

Dam, H. van

1982-01-01

A survey is given of the main incentives for power reactor noise research and the differences and similarities of noise in power and zero power systems are touched on. The basic characteristics of the adjoint method in reactor noise theory are treated. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurement of the reactor transfer function, which is demonstrated by results from measurements on a BWR in the Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

Interpretation of incore noise measurements in BWR's

International Nuclear Information System (INIS)

Dam, H. van

1983-01-01

A survey is given of the main incentives for power reactor noise research, and the differences and similarities of noise in power and zero power systems are shown. After a short outline of historical developments the basic characteristics of the adjoint method in reactor noise theory are dealt with. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies, which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurements on a BWR in The Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

Diseño e integración de algoritmos criptográficos en sistemas empotrados sobre FPGA

Directory of Open Access Journals (Sweden)

Alejandro Cabrera Aldaya

2013-10-01

Full Text Available En este trabajo se integran implementaciones hardware de algoritmos criptográficos a la biblioteca OpenSSL la cual es utilizadapor aplicaciones sobre el sistema operativo Linux para asegurar redes TCP/IP. Los algoritmos implementados son el AES y las funciones resumen SHA-1 y SHA-256. Estos algoritmos son implementados como coprocesadores del procesador MicroBlaze utilizando interfaces FSL para el intercambio de datos entre ellos. Estos coprocesadores son integrados dentro de la biblioteca OpenSSL considerando la naturaleza multitarea del sistema operativo Linux, por lo que se selecciona un mecanismo de sincronización para controlar el acceso a estos dispositivos. Además son presentados los resultados de velocidad alcanzados por los coprocesadores integrados en la biblioteca utilizando la herramienta speed de la misma. Finalmente es presentado el impacto de estos coprocesadores en la velocidad de transmisión a través de una red privada virtual utilizando la herramienta OpenVPN.

The HAMBO BWR simulator of HAMMLAB

International Nuclear Information System (INIS)

Karlsson, Tommy; Jokstad, Haakon; Meyer, Brita D.; Nihlwing, Christer; Norrman, Sixten; Puska, Eija Karita; Raussi, Pekka; Tiihonen, Olli

2001-02-01

Modernisation of control rooms of the nuclear power plants has been a major issue in Sweden and Finland the last few years, and this will continue in the years to come. As an aid in the process of introducing new technology into the control rooms, the benefit of having an experimental simulator where proto typing of solutions can be performed, has been emphasised by many plants. With this as a basis, the BWR plants in Sweden and Finland decided to fund, in co-operation with the Halden Project, an experimental BWR simulator based on the Forsmark 3 plant in Sweden. The BWR simulator development project was initiated in January 1998. VTT Energy in Finland developed the simulator models with the aid of their APROS tool, while the operator interface was developed by the Halden Project. The simulator was thoroughly tested by experienced HRP personnel and professional Forsmark 3 operators, and accepted by the BWR utilities in June 2000. The acceptance tests consisted of 19 well-defined transients, as well as the running of the simulator from full power down to cold shutdown and back up again with the use of plant procedures. This report describes the HAMBO simulator, with its simulator models, the operator interface, and the underlying hardware and software infrastructure. The tools used for developing the simulator, APROS, Picasso-3 and the Integration Platform, are also briefly described. The acceptance tests are described, and examples of the results are presented, to illustrate the level of validation of the simulator. The report concludes with an indication of the short-term usage of the simulator. (Author)

Development of next BWR plant

International Nuclear Information System (INIS)

Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke

1995-01-01

It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.)

Development of next BWR plant

Energy Technology Data Exchange (ETDEWEB)

Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

1995-04-01

It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.).

Transmutation of minor actinide using thorium fueled BWR core

International Nuclear Information System (INIS)

Susilo, Jati

2002-01-01

One of the methods to conduct transmutation of minor actinide is the use of BWR with thorium fuel. Thorium fuel has a specific behaviour of producing a little secondary minor actinides. Transmutation of minor actinide is done by loading it in the BWR with thorium fuel through two methods, namely close recycle and accumulation recycle. The calculation of minor actinide composition produced, weigh of minor actinide transmuted, and percentage of reminder transmutation was carried SRAC. The calculations were done to equivalent cell modeling from one fuel rod of BWR. The results show that minor actinide transmutation is more effective using thorium fuel than uranium fuel, through both close recycle and accumulation recycle. Minor actinide transmutation weight show that the same value for those recycle for 5th recycle. And most of all minor actinide produced from 5 unit BWR uranium fuel can transmuted in the 6 t h of close recycle. And, the minimal value of excess reactivity of the core is 12,15 % Δk/k, that is possible value for core operation

Decay ratio studies in BWR and PWR using wavelet

International Nuclear Information System (INIS)

Ciftcioglu, Oe.

1996-10-01

The on-line stability of BWR and PWR is studied using the neutron noise signals as the fluctuations reflect the dynamic characteristics of the reactor. Using appropriate signal modeling for time domain analysis of noise signals, the stability parameters can be directly obtained from the system impulse response. Here in particular for BWR, an important stability parameter is the decay ratio (DR) of the impulse response. The time series analysis involves the autoregressive modeling of the neutron detector signal. The DR determination is strongly effected by the low frequency behaviour since the transfer function characteristic tends to be a third order system rather than a second order system for a BWR. In a PWR low frequency behaviour is modified by the Boron concentration. As a result of these phenomena there are difficulties in the consistent determination of the DR oscillations. The enhancement of the consistency of this DR estimation is obtained by wavelet transform using actual power plant data from BWR and PWR. A comparative study of the Restimation with and without wavelets are presented. (orig.)

BWR radiation exposure--experience and projection

International Nuclear Information System (INIS)

Falk, C.F.; Wilkinson, C.D.; Hollander, W.R.

1979-01-01

The BWR/6 Mark III radiation exposures are projected to be about half of those of current average operating experience of 725 man-rem. These projections are said to be realistic and based on current achievements and not on promises of future development. The several BWRs operating with low primary system radiation levels are positive evidence that radiation sources can be reduced. Improvements have been made in reducing the maintenance times for the BWR/6, and further improvements can be made by further attention to cost-effective plant arrangement and layout during detail design to improve accessibility and maintainability of each system and component

The BWR Hybrid 4 control rod

International Nuclear Information System (INIS)

Gross, H.; Fuchs, H.P.; Lippert, H.J.; Dambietz, W.

1988-01-01

The service life of BWR control rods designed in the past has been unsatisfactory. The main reason was irradiation assisted stress corrosion cracking of B 4 C rods caused by external swelling of the B 4 C powder. By this reason KWU developed an improved BWR control rod (Hybrid 4 control rod) with extended service life and increased control rod worth. It also allows the procedure for replacing and rearranging fuel assemblies to be considerably simplified. A complete set of Hydbrid 4 control rods is expected to last throughout the service life of a plant (assumption: ca. 40 years) if an appropriate control rod reshuffling management program is used. (orig.)

An overview of the BWR ECCS strainer blockage issues

International Nuclear Information System (INIS)

Serkiz, A.W.; Marshall, M.L. Jr.; Elliott, R.

1996-01-01

This Paper provides a brief overview of actions taken in the mid 1980s to resolve Unresolved Safety Issue (USI) A-43, open-quotes Containment Emergency Sump Performance,close quotes and their relationship to the BWR strainer blockage issue; the importance of insights gained from the Barseback-2 (a Swedish BWR) incident in 1992 and from ECCS strainer testing and inspections at the Perry nuclear power plant in 1992 and 1993; an analysis of an US BWR/4 with a Mark I containment; an international community sharing of knowledge relevant to ECCS strainer blockage, additional experimental programs; and identification of actions needed to resolve the strainer blockage issue and the status of such efforts

Algoritmos para problemas de corte e empacotamento

OpenAIRE

Thiago Alves de Queiroz

2010-01-01

Resumo: Problemas de Corte e Empacotamento são, em sua maioria, NP-difíceis e não existem algoritmos exatos de tempo polinomial para tais se for considerado P ¿ NP. Aplicações práticas envolvendo estes problemas incluem a alocação de recursos para computadores; o corte de chapas de ferro, de madeira, de vidro, de alumínio, peças em couro, etc.; a estocagem de objetos; e, o carregamento de objetos dentro de contêineres ou caminhões-baú. Nesta tese investigamos problemas de Corte e Empacotament...

Algoritmo de lanczos na variedade de grassmann

OpenAIRE

Lopes, Ana Paula Ferreira Fernandes; Viamonte, Ana Júlia; Pascoal, António José

2008-01-01

O problema do cálculo de valores próprios, vectores próprios e subespaços invariantes está presente em áreas tão diversas como Engenharia, Física, Ciências de Computação e Matemática. Considerando a importância deste problema em tantas aplicações práticas, não é de surpreender que tenha sido e continue a ser objecto de intensa investigação, dando corpo a uma literatura muito vasta. Desenvolvemos um novo algoritmo de Lanczos na variedade de Grassmann. Este trabalho surgiu na ...

BWR type nuclear reactors

International Nuclear Information System (INIS)

Yamamoto, Toru.

1987-01-01

Purpose: To obtain reactor core characteristics with less changes in the excess reactivity due to fuel burnup even when the operation period varies. Constitution: In a BWR type reactor where fuel assemblies containing fuel rods incorporated with burnable poisons are arranged, the fuel assemblies are grouped into first fuel assemblies and second fuel assemblies. Then, the number of fuel rods incorporated with burnable poisons within the first fuel assemblies is made greater than that of the second fuel rods, while the concentration of the burnable poisons in the fuel rods incorporated with the burnable poisons in the first fuel assemblies is made lower than that of the fuel rods incorporated with the burnable poisons in the second fuel assemblies. In the BWR type reactor constituted in this way, the reactor core characteristics can be improved by changing the ratio between the first fuel assemblies and the second fuel assemblies charged to the reactor core, thereby decreasing the changes in the burnup of the excess reactivity. (Kamimura, M.)

Desarrollo de un sistema basado en la visión artificial para el reconocimiento de placas vehiculares

OpenAIRE

Suárez Pino, José Antonio; Suárez Pino, José Antonio; Suárez Pino, José Antonio

2011-01-01

El presente trabajo describe el desarrollo del sistema de reconocimiento de placas vehiculares basado en visión artificial. En primer lugar, la placa es detectada dentro de la imagen digital usando las características de ancho, altura y área de los objetos que hay en la imagen. Después de ubicar la placa, los caracteres son extraídos de la imagen uno a uno. Finalmente, cada carácter es reconocido usando el algoritmo de redes neuronales artificial. Las pruebas de reconocimiento se realizaron t...

Operator training simulator for BWR nuclear power plant

International Nuclear Information System (INIS)

Watanabe, Tadasu

1988-01-01

For the operation management of nuclear power stations with high reliability and safety, the role played by operators is very important. The effort of improving the man-machine interface in the central control rooms of nuclear power stations is energetically advanced, but the importance of the role of operators does not change. For the training of the operators of nuclear power stations, simulators have been used from the early stage. As the simulator facilities for operator training, there are the full scope simulator simulating faithfully the central control room of an actual plant and the small simulator mainly aiming at learning the plant functions. For BWR nuclear power stations, two full scope simulators are installed in the BWR Operator Training Center, and the training has been carried out since 1974. The plant function learning simulators have been installed in respective electric power companies as the education and training facilities in the companies. The role of simulators in operator training, the BTC No.1 simulator of a BWR-4 of 780 MWe and the BTC No.2 simulator of a BWR-5 of 1,100 MWe, plant function learning simulators, and the design of the BTC No.2 simulator and plant function learning simulators are reported. (K.I.)

FIST small break accident analysis with BWR TRACBO2-pretest predictions

International Nuclear Information System (INIS)

Alamgir, M.; Sutherland, W.A.

1984-01-01

The BWR Full Integral Simulation Test (FIST) program includes experimental simulation and analytical evaluation of BWR thermal-hydraulic phenomena during transient events. One such event is a small size break in the suction line of one of the recirculation pumps. The results from a test simulating this transient in the FIST facility are compared with a system analysis using the Transient Reactor Analysis Code (TRACB02). This comparison demonstrates BWR-TRAC capability for small break analyses and provides detailed understanding of the phenomena

Algoritmos genéticos para la resolución de problemas de Programación por Metas Entera. Aplicación a la Economía de la Educación.

Directory of Open Access Journals (Sweden)

Caballero Fernández, Rafael

2002-01-01

Full Text Available En este trabajo se presenta un algoritmo genético para la resolución de problemas de Programación por Metas Entera. Este tipo de problemas presentan en general un gran número de dificultades para su resolución utilizando algoritmos tradicionales de Programación Entera, siendo en la mayoría de los casos de problemas reales demasiado costosa computacionalmente para afrontarla con garantías. Sin embargo, se mostrará como este nuevo tipo de algoritmos, los algoritmos genéticos, permiten resolver eficientemente problemas de este tipo con un coste computacional reducido. Como ejemplo, se resuelve en este trabajo un problema Economía de la Educación con un modelo de Programación por Metas Entera aplicando un algoritmo genético y un algoritmo tradicional. Para estas resoluciones se analiza el coste computacional de ambos tipos de resolución para poner de manifiesto las ventajas que puede suponer un algoritmo genético para la resolución de problemas reales complejos.

Algoritmo de um teste adaptativo informatizado com base na teoria da resposta ao item para a estimação da usabilidade de sites de e-commerce

Directory of Open Access Journals (Sweden)

Fernando de Jesus Moreira Junior

2013-09-01

Full Text Available O presente artigo propõe um algoritmo de um teste adaptativo informatizado baseado na teoria da resposta ao item, desenvolvido para estimar o grau de usabilidade de sites de e-commerce. Cinco algoritmos baseados no critério da máxima informação foram desenvolvidos e testados via simulação. O algoritmo com o melhor desempenho foi aplicado nos dados reais de 361 sites de e-commerce. Os resultados mostraram que o algoritmo desenvolvido consegue obter uma boa estimativa para o grau de usabilidade de sites de e-commerce com a aplicação de 13 itens.

Algoritmo para el tratamiento mínimamente invasivo de la coledocolitiasis

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Javier Ernesto Barreras González

Full Text Available La coledocolitiasis se presenta en más del 10 % de los pacientes jóvenes sometidos a la colecistectomía laparoscópica y se incrementa por encima de 25 % en los pacientes mayores de 60 años. Una vez que se identifica el cálculo en la vía biliar principal (VBP, esto puede convertirse en un dilema terapéutico. En la actualidad, diferentes algoritmos se han planteado y están establecidos, pero las publicaciones internacionales continúan redefiniendo las indicaciones de la colangiopancreatografía retrógrada endoscópica sobre la exploración laparoscópica de la vía biliar principal y viceversa. El objetivo de este estudio fue presentar el algoritmo para el tratamiento de la coledocolitiasis del Centro Nacional de Cirugía de Mínimo Acceso de La Habana. Se realizó una amplia revisión bibliográfica acerca del manejo actual de la coledocolitiasis con vesícula in situ y se evaluaron los resultados de dichas técnicas en el Centro Nacional de Cirugía de Mínimo Acceso. Las opciones para el manejo de la coledocolitiasis incluyen la exploración transcística de la vía biliar principal, la exploración laparoscópica de la vía biliar principal mediante coledocotomía, la colangiopancreatografía retrógrada endoscópica preoperatoria e intraoperatoria con esfinterotomía. Se presenta el algoritmo para el tratamiento mínimamente invasivo de la coledocolitiasis. El manejo de esta enfermedad en la actualidad depende de la experiencia de cada equipo quirúrgico y de su disponibilidad de acceso a los avances tecnológicos.

Algoritmos paralelos numéricos para a resolução de sistemas de equações lineares

OpenAIRE

Maria Paula Goncalves Fachin

1989-01-01

Apresentamos as características de algoritmos para obter a solução de sistemas de equações lineares, utilizando computadores com arquitetura paralela e explorando os recursos de paralelismo neles embutidos. São descritos os modelos de computadores paralelos e os procedimentos para a criação de algoritmos paralelos. The characteristics of algorithms to solve systems of linear equations, using parallel computers, are presented. The abstract models of these computers are described, as well as...

VALIDACIÓN DE UN ALGORITMO HÍBRIDO DEL PSO CON EL MÉTODO SIMPLEX Y DE TOPOLOGÍA DE EVOLUCIÓN PARAMÉTRICA

OpenAIRE

RODRIGO CORREA; OSCAR BEGAMBRE; JULIO C. CARRILLO E.

2011-01-01

Este artículo describe algunos de los aspectos más importantes relacionados con la experimentación numérica de un híbrido del algoritmo PSO (Particle Swarm Optimization) con el tradicional método simplex modificado de Nelder-Mead. El híbrido de estas dos técnicas de optimización sin restricciones se realizó con una topología que permite optimizar en cada iteración los parámetros del algoritmo PSO utilizando el método simplex modificado. Se realizaron experimentos numéricos con este algoritmo ...

The JAERI code system for evaluation of BWR ECCS performance

International Nuclear Information System (INIS)

Kohsaka, Atsuo; Akimoto, Masayuki; Asahi, Yoshiro; Abe, Kiyoharu; Muramatsu, Ken; Araya, Fumimasa; Sato, Kazuo

1982-12-01

Development of respective computer code system of BWR and PWR for evaluation of ECCS has been conducted since 1973 considering the differences of the reactor cooling system, core structure and ECCS. The first version of the BWR code system, of which developmental work started earlier than that of the PWR, has been completed. The BWR code system is designed to provide computational tools to analyze all phases of LOCAs and to evaluate the performance of the ECCS including an ''Evaluation Model (EM)'' feature in compliance with the requirements of the current Japanese Evaluation Guideline of ECCS. The BWR code system could be used for licensing purpose, i.e. for ECCS performance evaluation or audit calculations to cross-examine the methods and results of applicants or vendors. The BWR code system presented in this report comprises several computer codes, each of which analyzes a particular phase of a LOCA or a system blowdown depending on a range of LOCAs, i.e. large and small breaks in a variety of locations in the reactor system. The system includes ALARM-B1, HYDY-B1 and THYDE-B1 for analysis of the system blowdown for various break sizes, THYDE-B-REFLOOD for analysis of the reflood phase and SCORCH-B2 for the calculation of the fuel assembl hot plane temperature. When the multiple codes are used to analyze a broad range of LOCA as stated above, it is very important to evaluate the adequacy and consistency between the codes used to cover an entire break spectrum. The system consistency together with the system performance are discussed for a large commercial BWR. (author)

Report on the BWR owners group radiation protection/ALARA Committee

International Nuclear Information System (INIS)

Aldrich, L.R.

1995-01-01

Radiation protection programs at U.S. boiling water reactor (BWR) stations have evolved during the 1980s and early 1990s from a regulatory adherence-based endeavor to a proactive, risk-based radiation protection and prevention mission. The objectives are no longer to merely monitor and document exposure to radiation and radioactive materials. The focus of the current programs is the optimization of radiation protection of occupational workers consistent with the purpose of producing cost-effective electric power. The newly revised 10 CFR 20 defines the term ALARA (as low as reasonably achievable) to take into account the state of technology, the economics of improvements in relation to the state of the technology, and the benefits to the public health and safety. The BWR Owners Group (BWROG) initially formed the Radiation Protection/ALARA Committee in January 1990 to evaluate methods of reducing occupational radiation exposure during refueling outages. Currently, twenty U.S. BWR owner/operators (representing 36 of the operational 37 domestic BWR units), as well as three foreign BWR operators (associate members), have broadened the scope to promote information exchange between BWR radiation protection professionals and develop good practices which will affect optimization of their radiation protection programs. In search of excellence and the challenge of becoming open-quotes World Classclose quotes performers in radiation protection, the BWROG Radiation Protection/ALARA Committee has recently accepted a role in assisting the member utilities in improving radiation protection performance in a cost-effective manner. This paper will summarize the recent activities of this Committee undertaken to execute their role of exchanging information in pursuit of optimizing the improvement of their collective radiation protection performance

Report on the BWR owners group radiation protection/ALARA Committee

Energy Technology Data Exchange (ETDEWEB)

Aldrich, L.R. [Commonwealth Edison Co., Downers Grove, IL (United States)

1995-03-01

Radiation protection programs at U.S. boiling water reactor (BWR) stations have evolved during the 1980s and early 1990s from a regulatory adherence-based endeavor to a proactive, risk-based radiation protection and prevention mission. The objectives are no longer to merely monitor and document exposure to radiation and radioactive materials. The focus of the current programs is the optimization of radiation protection of occupational workers consistent with the purpose of producing cost-effective electric power. The newly revised 10 CFR 20 defines the term ALARA (as low as reasonably achievable) to take into account the state of technology, the economics of improvements in relation to the state of the technology, and the benefits to the public health and safety. The BWR Owners Group (BWROG) initially formed the Radiation Protection/ALARA Committee in January 1990 to evaluate methods of reducing occupational radiation exposure during refueling outages. Currently, twenty U.S. BWR owner/operators (representing 36 of the operational 37 domestic BWR units), as well as three foreign BWR operators (associate members), have broadened the scope to promote information exchange between BWR radiation protection professionals and develop good practices which will affect optimization of their radiation protection programs. In search of excellence and the challenge of becoming {open_quotes}World Class{close_quotes} performers in radiation protection, the BWROG Radiation Protection/ALARA Committee has recently accepted a role in assisting the member utilities in improving radiation protection performance in a cost-effective manner. This paper will summarize the recent activities of this Committee undertaken to execute their role of exchanging information in pursuit of optimizing the improvement of their collective radiation protection performance.

Advanced methods for BWR transient and stability analysis

Energy Technology Data Exchange (ETDEWEB)

Schmidt, A; Wehle, F; Opel, S; Velten, R [AREVA, AREVA NP, Erlangen (Germany)

2008-07-01

The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

Advanced methods for BWR transient and stability analysis

International Nuclear Information System (INIS)

Schmidt, A.; Wehle, F.; Opel, S.; Velten, R.

2008-01-01

The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

Aplicativo para análise comparativa do comportamento de algoritmos de ordenação

Directory of Open Access Journals (Sweden)

João Paulo Folador

2014-11-01

Full Text Available Este artigo apresenta uma ferramenta prática de auxílio pedagógico para demonstrar o comportamento de alguns algoritmos de ordenação. Nesse contexto, foram estudados e confrontados os algoritmos: bubble sort, merge sort, quick sort e o shell sort, de modo a ordenar uma matriz de elementos variáveis. Na aplicação desenvolvida é permitido ao usuário configurar a quantidade de elementos da matriz a serem ordenados, quais métodos serão utilizados, a quantidade de ciclos somados e, o resultado final, é mostrado por meio de um gráfico, comparando os métodos aplicados. Ainda, como saída do processo de ordenação, são mostrados: o tempo médio de execução, o desvio padrão, o tempo mínimo e o tempo máximo gasto para ordenar os elementos da matriz. Desse modo, essa aplicação de apoio didático torna-se prática para testes feitos pelos alunos nas disciplinas que envolvem o estudo de algoritmos de ordenação facilitando, assim, o processo de aprendizagem.

Study of behavior on bonding and failure mode of pressurized and doped BWR fuel rod

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki

1992-03-01

The study of transient behavior on the bonding and the failure mode was made using the pressurized/doped 8 x 8 BWR type fuel rod. The dopant was mullite minerals consisted mainly of silicon and aluminum up to 1.5 w/o. Pressurization of the fuel rod with pure helium was made to the magnitude about 0.6 MPa. As a reference, the non-pressurized/non-doped 8 x 8 BWR fuel rod and the pressurized/7 x 7 BWR fuel rod up to 0.6 MPa were prepared. Magnitude of energy deposition given to the tested fuel rods was 248, 253, and 269 cal/g·fuel, respectively. Obtained results from the pulse irradiation in NSRR are as follows. (1) It was found from the experiment that alternation of the fuel design by the adoption of pressurization up to 0.6 MPa and the use of wider gap up to 0.38 mm could avoid the dopant BWR fuel from the overall bonding. The failure mode of the present dopant fuel was revealed to be the melt combined with rupture. (2) The time of fuel failure of the pressurized/doped 8 x 8 BWR fuel defected by the melt/rupture mode is of order of two times shorter than that of the pressurized/ 7 x 7 BWR defected by the rupture mode. Failure threshold of the pressurized/doped 8 x 8 BWR BWR tended to be lower than that of non-pressurized/non-doped 8 x 8 BWR one. Cracked area of the pressurized/doped 8 x 8 BWR was more wider and magnitude of oxidation at the place is relatively larger than the other tested fuels. (3) Failure mode of the non-pressurized/ 8 x 8 BWR fuel rod was the melt/brittle accompanied with a significant bonding at failed location. While, failure mode of the pressurized/ 7 x 7 BWR fuel rod was the cladding rupture accompanied with a large ballooning. No bonding at failed location of the latter was observed. (author)

Assessment of the Prony's method for BWR stability analysis

International Nuclear Information System (INIS)

Ortiz-Villafuerte, Javier; Castillo-Duran, Rogelio; Palacios-Hernandez, Javier C.

2011-01-01

Highlights: → This paper describes a method to determine the degree of stability of a BWR. → Performance comparison between Prony's and common AR techniques is presented. → Benchmark data and actual BWR transient data are used for comparison. → DR and f results are presented and discussed. → The Prony's method is shown to be a robust technique for BWR stability. - Abstract: It is known that Boiling Water Reactors are susceptible to present power oscillations in regions of high power and low coolant flow, in the power-flow operational map. It is possible to fall in one of such instability regions during reactor startup, since both power and coolant flow are being increased but not proportionally. One other possibility for falling into those areas is the occurrence of a trip of recirculation pumps. Stability monitoring in such cases can be difficult, because the amount or quality of power signal data required for calculation of the stability key parameters may not be enough to provide reliable results in an adequate time range. In this work, the Prony's Method is presented as one complementary alternative to determine the degree of stability of a BWR, through time series data. This analysis method can provide information about decay ratio and oscillation frequency from power signals obtained during transient events. However, so far not many applications in Boiling Water Reactors operation have been reported and supported to establish the scope of using such analysis for actual transient events. This work presents first a comparison of decay ratio and frequency oscillation results obtained by Prony's method and those results obtained by the participants of the Forsmark 1 and 2 Boiling Water Reactor Stability Benchmark using diverse techniques. Then, a comparison of decay ratio and frequency oscillation results is performed for four real BWR transient event data, using Prony's method and two other techniques based on an autoregressive modeling. The four

MULTIOBJECTIVE EVOLUTIONARY ALGORITHMS APPLIED TO MICROSTRIP ANTENNAS DESIGN ALGORITMOS EVOLUTIVOS MULTIOBJETIVO APLICADOS A LOS PROYECTOS DE ANTENAS MICROSTRIP

Directory of Open Access Journals (Sweden)

Juliano Rodrigues Brianeze

2009-12-01

Full Text Available This work presents three of the main evolutionary algorithms: Genetic Algorithm, Evolution Strategy and Evolutionary Programming, applied to microstrip antennas design. Efficiency tests were performed, considering the analysis of key physical and geometrical parameters, evolution type, numerical random generators effects, evolution operators and selection criteria. These algorithms were validated through design of microstrip antennas based on the Resonant Cavity Method, and allow multiobjective optimizations, considering bandwidth, standing wave ratio and relative material permittivity. The optimal results obtained with these optimization processes, were confirmed by CST Microwave Studio commercial package.Este trabajo presenta tres de los principales algoritmos evolutivos: Algoritmo Genético, Estrategia Evolutiva y Programación Evolutiva, aplicados al diseño de antenas de microlíneas (microstrip. Se realizaron pruebas de eficiencia de los algoritmos, considerando el análisis de los parámetros físicos y geométricos, tipo de evolución, efecto de generación de números aleatorios, operadores evolutivos y los criterios de selección. Estos algoritmos fueron validados a través del diseño de antenas de microlíneas basado en el Método de Cavidades Resonantes y permiten optimizaciones multiobjetivo, considerando ancho de banda, razón de onda estacionaria y permitividad relativa del dieléctrico. Los resultados óptimos obtenidos fueron confirmados a través del software comercial CST Microwave Studio.

Analysis of multidimensional and countercurrent effects in a BWR loss-of-coolant accident

International Nuclear Information System (INIS)

Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

1991-01-01

The presence of parallel enclosed channels in a boiling water reactor (BWR) provides opportunities for multiple flow regimes in cocurrent and countercurrent flow under loss-of-coolant accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the steam sector test facility (SSFT), which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests an integral transients with vessel vlowdown and refill were performed. The presence of multidimensional and parallel-channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

ALGORITMO PARA EL APRENDIZAJE DE REGLAS DE CLASIFICACION BASADO EN LA TEORÍA DE LOS CONJUNTOS APROXIMADOS EXTENDIDA

Directory of Open Access Journals (Sweden)

YAIMA FILIBERTO

2011-01-01

Full Text Available Los conjuntos aproximados han demostrado ser efectivos para desarrollar técnicas de aprendizaje automático, entre ellos métodos para el descubrimiento de reglas de clasificación. En este trabajo se presenta un algoritmo para generar reglas de clasificación basado en relaciones de similaridad, lo que permite que sea aplicable en casos donde los rasgos tienen dominio discreto o continuo. Los resultados experimentales muestran un desempeño satisfactorio en comparación con otros algoritmos conocidos como C4.5 y MODLEM.

Algoritmo genético permutacional para el despliegue y la planificación de sistemas de tiempo real distribuidos

Directory of Open Access Journals (Sweden)

Ekain Azketa

2013-07-01

Full Text Available Resumen: El despliegue y la planificación de tareas y mensajes en sistemas de tiempo real distribuidos son problemas NP-difíciles (NP- hard, por lo que no existen métodos óptimos para solucionarlos en tiempo polinómico. En consecuencia, estos problemas son adecuados para abordarse mediante algoritmos genéricos de búsqueda y optimización. En este artículo se propone un algoritmo genético multiobjetivo basado en una codificación permutacional de las soluciones para abordar el despliegue y la planificación de sistemas de tiempo real distribuidos. Además de desplegar tareas en computadores y de planificar tareas y mensajes, este algoritmo puede minimizar el número de computadores utilizados, la cantidad de recursos computacionales y de comunicaciones empleados y el tiempo de respuesta de peor caso medio de las aplicaciones. Los resultados experimentales muestran que este algoritmo genético permutacional puede desplegar y planificar sistemas de tiempo real distribuidos de forma satisfactoria y en tiempos razonables. Abstract: The deployment and scheduling of tasks and messages in distributed real-time systems are NP-hard problems, so there are no optimal methods to solve them in polynomial time. Consequently, these problems are suitable to be approached with generic search and optimisation algorithms. In this paper we propose a multi-objective genetic algorithm based on a permutational solution encoding for the deployment and scheduling of distributed real-time systems. Besides deploying and scheduling tasks and messages, the algorithm can minimize the number of the used computers, the utilization of computing and networking resources and the average worst-case response times of the applications. The experiments show that this genetic algorithm can successfully synthesize complex distributed real-time systems in reasonable times. Palabras clave: Sistemas de tiempo real, Algoritmos de planificaci

An A BWR demonstration simulator for training and developing technical staff

International Nuclear Information System (INIS)

Powers, J.; Yonezawa, H.; Aoyagi, Y.; Kataoka, K.

2015-09-01

The US-Advanced Boiling Water Reactor (A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. Toshiba has developed a Demonstration Simulator of the A BWR control room that provides a realistic experience for training and education on BWR principles and operations fundamentals. The Demonstration Simulator is located in the Toshiba America Nuclear Energy (Tane) office in Charlotte, North Carolina and is composed of standard office computer equipment set up in a specific arrangement that is representative of the layout of an A BWR control room. The Demonstration Simulator is not intended for licensed operator training, but can provide a framework for encouraging entry level technically oriented nuclear workers to enter the operations field; strengthening the linkage between university energy field curricula and real-life application of theory; and, improving understanding of integrated plant operations for developing station technical staff. This paper describes the A BWR Demonstration Simulator and its applications for training and educating future nuclear workers. (Author)

An A BWR demonstration simulator for training and developing technical staff

Energy Technology Data Exchange (ETDEWEB)

Powers, J. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Yonezawa, H.; Aoyagi, Y.; Kataoka, K., E-mail: jim.powers@toshiba.com [Toshiba Corporation, Kawasaki, Kanagawa (Japan)

2015-09-15

The US-Advanced Boiling Water Reactor (A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. Toshiba has developed a Demonstration Simulator of the A BWR control room that provides a realistic experience for training and education on BWR principles and operations fundamentals. The Demonstration Simulator is located in the Toshiba America Nuclear Energy (Tane) office in Charlotte, North Carolina and is composed of standard office computer equipment set up in a specific arrangement that is representative of the layout of an A BWR control room. The Demonstration Simulator is not intended for licensed operator training, but can provide a framework for encouraging entry level technically oriented nuclear workers to enter the operations field; strengthening the linkage between university energy field curricula and real-life application of theory; and, improving understanding of integrated plant operations for developing station technical staff. This paper describes the A BWR Demonstration Simulator and its applications for training and educating future nuclear workers. (Author)

Power oscillations in BWR reactors

International Nuclear Information System (INIS)

Espinosa P, G.

2002-01-01

One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

Scaling and uncertainty in BWR instability problems

International Nuclear Information System (INIS)

Di Auria, F.; Pellicoro, V.

1995-01-01

This paper deals with a critical review of activities, performed at the DCMN of Pisa University, in relation to the thermo-hydraulic oscillations in two-phase systems. Stability analyses, including model development and achievement of experimental data, are generally performed for BWRs in order to achieve the following objectives: to reach a common understanding in relation to the predictive capabilities of system codes and to the influence of various parameters on the instability; to establish a data base for the qualification of the analytical tools already or becoming available; to set-up qualified tools (code/models + nodalization + user assumption) suitable for predicting the unstable behaviour of the nuclear plants of interest (current BWR, SBWR, ABWR and RBMK). These considerations have been the basis for the following researches: 1) proposal of the Boiling Instability Program (BIP) (1) 2) evaluation of stability tests in PIPER-ONE apparatus (2) 3) coupled thermal-hydraulic and neutronic instabilities in the LaSalle-2 BWR plant (3) 4) participation to the NEA-OECD BWR Benchmark (4) The RELAP/MOD2 and RELAP5/MOD3 codes have been used. (author)

Tritium in liquid phase in a BWR-5 like Laguna Verde; Tritio en fase liquida en un BWR-5 como Laguna Verde

Energy Technology Data Exchange (ETDEWEB)

Tijerina S, F.; Vargas A, A.; Cardenas J, J., E-mail: francisco.tijerina@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Carretera Veracruz-Medellin Km 7.5, Veracruz (Mexico)

2011-11-15

In boiling water reactors (BWR), the tritium (H{sub 3}) takes place mainly as a result of ternary fissions in the nuclear reactors, of those which 75% are in gaseous form and 25% in liquid form. In the liquid phase, the tritium is transported to the pipes of the primary coolant toward condensed tanks or tanks of drainage excesses of radioactive equipment, located in external areas of a BWR, as well as to the processes of radioactive wastes to be able to be directed to the liquid effluents. For that reason, is necessary to know the possible routes of the transport and processes of the tritium in a BWR to control this radioisotope in the site of the event of leaks in equipment s and buried pipes, avoiding that emigrates toward underground flowing and an impact to the environment and to the people in general. (Author)

Material operating behaviour of ABB BWR control rods

International Nuclear Information System (INIS)

Rebensdorff, B.; Bart, G.

2000-01-01

The BWR control rods made by ABB use boron carbide (B 4 C and hafnium as absorber material within a cladding of stainless steel. The general behaviour under operation has proven to be very good. ABB and many of their control rod customers have performed extensive inspection programs of control rod behaviour. However, due to changes in the material properties under fast and thermal neutron irradiation defects may occur in the control rods at high neutron fluences. Examinations of irradiated control rod materials have been performed in hot cell laboratories. The examinations have revealed the defect mechanism Irradiation Assisted Stress Corrosion Cracking (IASCC) to appear in the stainless steel cladding. For IASCC to occur three factors have to act simultaneously. Stress, material sensitization and an oxidising environment. Stress may be obtained from boron carbide swelling due to irradiation. Stainless steel may be sensitized to intergranular stress corrosion cracking under irradiation. Normally the reactor environment in a BWR is oxidising. The presentation focuses on findings from hot cell laboratory work on irradiated ABB BWR control rods and studies of irradiated control rod materials in the hot cells at PSI. Apart from physical, mechanical and microstructural examinations, isotope analyses were performed to describe the local isotopic burnup of boron. Consequences (such as possible B 4 C washout) of a under operation in a ABB BWR, after the occurrence of a crack is discussed based on neutron radiographic examinations of control rods operated with cracks. (author)

General Electric's training program for BWR chemists

International Nuclear Information System (INIS)

Osborn, R.N.; Lim, W.

1981-01-01

This paper describes the development and implementation of the General Electric boiling water reactor chemistry training program from 1959 to the present. The original intention of this program was to provide practical hands on type training in radiochemistry to BWR chemistry supervisors with fossil station experience. This emphasis on radiochemistry has not changed through the years, but the training has expanded to include the high purity water chemistry of the BWR and has been modified to include new commission requirements, engineering developments and advanced instrumentation. Student and instructor qualifications are discussed and a description of the spin off courses for chemistry technicians and refresher training is presented

BWR SFAT, gross-defect verification of spent BWR fuel. Final report on Task FIN A563 on the Finnish Support Programme to IAEA Safeguards including BWR SFAT User Manual

International Nuclear Information System (INIS)

Tarvainen, M.; Paakkunainen, M.; Tiitta, A.; Sarparanta, K.

1994-04-01

A measurement instrument called Spent Fuel Attribute Tester, SFAT, has been designed, fabricated and taken into use by the IAEA in gross defect verification of spent BWR fuel assemblies. The equipment consists of an underwater measurement head connected with cables to a control unit on the bridge of the fuel handling machine as well as to a PMCA for measurement of the gamma spectra. The BWR SFAT is optimized for the AFR interim storage, TVO KPA-STORE, of the TVO Power Company in Olkiluoto, Finland. It has a shape and it is moved like a fuel assembly using the fuel handling machine. No fuel movements are needed. Spent fuel specific radiation from the fission product 137 Cs at the gamma-ray energy of 662 keV is detected above the assemblies in the storage rack using a NaI(Tl) detector. In the design and in licensing the requirements of the IAEA, operator and the safety authority have been taken into account. The BWR SFAT allows modifications for other LWR fuel types with minor changes. The work has been carried out under the task FIN A 563 of the Finnish Support Programme to IAEA Safeguards. (orig.) (9 refs., 22 figs.)

Reducing BWR O and M costs through on-line performance monitoring

International Nuclear Information System (INIS)

Jonas, T.; Gross, R.; Logback, F.; Josyula, R.

1995-01-01

Competition in the electric power industry has placed significant emphasis on reducing operating and maintenance (O and M) costs at nuclear facilities. Therefore, on-line performance monitoring to locate power losses for boiling water reactor (BWR) plants is creating tremendous interest. In addition, the ability to automate activities such as data collection, analysis, and reporting increases the efficiency of plant engineers and gives them more time to concentrate on solving plant efficiency problems. This capability is now available with a unique software product called GEBOPS. GE Nuclear Energy, in conjunction with Joint Venture partner Black and Veatch, has undertaken development of the General Electric/Black and Veatch On-line Performance System (GEBOPS), an on-line performance monitoring system for BWR plants. The experience and expertise of GE Nuclear Energy with BWR plants, coupled with the proven on-line monitoring software development experience and capability of Black and Veatch, provide the foundation for a unique product which addresses the needs of today's BWR plants

Peach Bottom transient analysis with BWR TRACB02

International Nuclear Information System (INIS)

Alamgir, M.; Sutherland, W.A.

1984-01-01

TRAC calculations have been performed for a Turbine Trip transient (TT1) in the Peach Bottom BWR power plant. This study is a part of the qualification of the BWR-TRAC code. The simulation is aimed at reproducing the observed thermal hydraulic behavior in a pressurization transient. Measured core power is an input to the calculation. Comparison with data show the code reasonably well predicts the generation and propagation of the pressure waves in the main steam line and associated pressurization of the reactor vessel following the closure of the turbine stop valve

Methyl Iodide Decomposition at BWR Conditions

International Nuclear Information System (INIS)

Pop, Mike; Bell, Merl

2012-09-01

Based on favourable results from short-term testing of methanol addition to an operating BWR plant, AREVA has performed numerous studies in support of necessary Engineering and Plant Safety Evaluations prior to extended injection of methanol. The current paper presents data from a study intended to provide further understanding of the decomposition of methyl iodide as it affects the assessment of methyl iodide formation with the application of methanol at BWR Plants. This paper describes the results of the decomposition testing under UV-C light at laboratory conditions and its effect on the subject methyl iodide production evaluation. The study as to the formation and decomposition of methyl iodide as it is effected by methanol addition is one phase of a larger AREVA effort to provide a generic plant Safety Evaluation prior to long-term methanol injection to an operating BWR. Other testing phases have investigated the compatibility of methanol with fuel construction materials, plant structural materials, plant consumable materials (i.e. elastomers and coatings), and ion exchange resins. Methyl iodide is known to be very unstable, typically preserved with copper metal or other stabilizing materials when produced and stored. It is even more unstable when exposed to light, heat, radiation, and water. Additionally, it is known that methyl iodide will decompose radiolytically, and that this effect may be simulated using ultra-violet radiation (UV-C) [2]. In the tests described in this paper, the use of a UV-C light source provides activation energy for the formation of methyl iodide. Thus is similar to the effect expected from Cherenkov radiation present in a reactor core after shutdown. Based on the testing described in this paper, it is concluded that injection of methanol at concentrations below 2.5 ppm in BWR applications to mitigate IGSCC of internals is inconsequential to the accident conditions postulated in the FSAR as they are related to methyl iodide formation

TRAC-BWR development

International Nuclear Information System (INIS)

Weaver, W.L.; Rouhani, S.Z.

1983-01-01

The TRAC-BD1/MOD1 code containing many new or improved models has been assembled and is undergoing developmental assessment and testing and should be available shortly. The preparation of the manual for this code version is underway and should be available to the USNRC and their designated contractors by April of 1984. Finally work is currently underway on a fast running version of TRAC-BWR which will contain a one-dimensional neutron kinetics model

BWR mechanics and materials technology update

International Nuclear Information System (INIS)

Kiss, E.

1983-01-01

This paper discusses technical results obtained from a variety of important programs underway at General Electric's Nuclear Engineering Division. The principal objective of these programs is to qualify and improve BWR product related technologies that fall broadly under the disciplines of Applied Mechanics and Materials Engineering. The paper identifies and deals with current technical issues that are of general importance to the LWR industry albeit the specific focus is directed to the development and qualification of analytical predictive methods and criteria, and improved materials for use in the design of the BWR. In this paper, specific results and accomplishments are summarized to provide a braod perspective of technology advances. Results are presented in sections which discuss: dynamic analysis and modeling; fatigue and fracture evaluation; materials engineering advances; and flow induced vibration. (orig.)

Water chemistry control practices and data of the European BWR fleet

International Nuclear Information System (INIS)

Stellwag, B.; Laendner, A.; Weiss, S.; Huettner, F.

2010-01-01

Nineteen BWR plants are in operation in Europe, nine built by ASEA Atom, six by Siemens KWU and four by General Electric. This paper gives an overview of water chemistry operation practices and parameters of the European BWR plants. General design characteristics of the plants are described. Chemistry control strategies and underlying water chemistry guidelines are summarized. Chemistry data are presented and discussed with regard to plant design characteristics. The paper is based on a contract of the European BWR Forum with AREVA on a chemistry sourcebook for member plants. The survey of chemistry data was conducted for the years 2002 to 2008. (author)

The BWR owners' group planning guide for life extension

International Nuclear Information System (INIS)

Smith, S.K.; Lehnert, D.F.; Locke, R.K.

1991-01-01

Extending the operating life of a commercial nuclear power plant has been shown to be economically beneficial to both the utility and the electric customer. As such, many utilities are planning and implementing plant life extension (PLEX) programs. A document has been developed which provides guidance to utilities in formulating a PLEX program plant for one or more boiling water reactor (BWR) plants. The guide has been developed by the BWR Owners' Group Plant Life Extension Committee. The principal bases for this guide were the BWR Pilot and Lead Plant Programs. These programs were used as models to develop the 'base plan' described in this guide. By formulating their program plant utilizing the base plan, utilities will be able to maximize the use of existing evaluations and results. The utility planner will build upon the base plan by adding any tasks or features that are unique to their programs. (author)

Simulation of decreasing reactor power level with BWR simulator

International Nuclear Information System (INIS)

Suwoto; Zuhair; Rivai, Abu Khalid

2002-01-01

Study on characteristic of BWR using Desktop PC Based Simulator Program was analysed. This simulator is more efficient and cheaper for analyzing of characteristic and dynamic respond than full scope simulator for decreasing power level of BW. Dynamic responses of BWR reactor was investigated during the power level reduction from 100% FP (Full Power) which is 3926 MWth to 0% FP with 25% steps and 1 % FP/sec rate. The overall results for core flow rate, reactor steam flow, feed-water flow and turbine-generator power show tendency proportional to reduction of reactor power. This results show that reactor power control in BWR could be done by control of re-circulation flow that alter the density of water used as coolant and moderator. Decreasing the re-circulation flow rate will decrease void density which has negative reactivity and also affect the position of control rods

Metallurgical factors that contribute to cracking in BWR piping

International Nuclear Information System (INIS)

Weeks, J.R.

1975-01-01

During the fall of 1974 and early winter of 1975, cracks have been discovered in the 4 in. bypass lines of several Boiling Water Reactors (BWR's) in the United States. Further, similar cracks were discovered at two BWR's in Japan during the same period. More recently, cracks have been discovered in the core spray piping and in a furnace-sensitized ''safe end'' and adjacent ''dutchman'' at the Dresden Nuclear Power Station, Unit No. 2. Although inspections at all other U.S. BWR's have not disclosed further instances of cracking in core spray piping, leaking cracks have been found in the core spray piping of two BWR's overseas. Metallurgical examinations of these cracks are not yet complete. The following observations have been made to date. All cracks (except those in the furnace-sensitized safe end and dutchman) occurred in seamless type 304 stainless steel piping or in elbows fabricated from such piping, in the outer heat affected zone of either field or shop welds, in lines isolated from the main primary coolant flow during full power operation, except for the not yet examined cracks in the Monticello bypass lines. The cracks are exclusively intergranular, and occur in metal that has been lightly sensitized by the welding process, with only intermittent grain boundary carbides. They developed in the areas of peak axial residual stresses from welding rather than in the most heavily sensitized areas. No fatigue striations have been found on the fracture surfaces. The evidence received to date strongly indicates that these cracks were caused by intergranular stress corrosion of weld-sensitized stainless steel by BWR water containing greater than 0.2 ppM oxygen. The possible role of fatigue or alternating stresses in this corrosion is not clear. Further, not all the cracks detected to date necessarily have occurred by the same mechanism

BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

International Nuclear Information System (INIS)

Garcia, S.E.; Giannelli, J.F.; Jarvis, M.L.

2010-01-01

The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

Energy Technology Data Exchange (ETDEWEB)

Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, M.L., E-mail: jgiannelli@finetech.com [Finetech, Inc., Parsippany, NJ (United States)

2010-07-01

The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

Advanced BWR core component designs and the implications for SFD analysis

International Nuclear Information System (INIS)

Ott, L.J.

1997-01-01

Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B 4 C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities

Implementación del Algoritmo Sünter-Clare en un Convertidor Matricial 3x3

Directory of Open Access Journals (Sweden)

Eliher A. Ortiz Colín

2017-10-01

Full Text Available Resumen: En este trabajo se presenta la implementación digital del algoritmo de modulación de Sünter-Clare, para un convertidor matricial de tres hilos de 7.5 kVA. Este algoritmo es usado para calcular los ciclos de trabajo en un convertidor matricial 3x3, con lo que se generan tensiones de salida con amplitud y frecuencia constantes. El algoritmo de modulación de Sünter-Clare recalcula los tiempos y trayectorias de conmutación cada período de muestreo, con el objetivo de compensar las variaciones de amplitud y de frecuencia de las tensiones de entrada. Este algoritmo se ejecuta en cada muestreo y está en función de la tensión trifásica de entrada y de la tensión de referencia de salida, resultando adecuado para controles en lazo cerrado, cuando las señales de amplitud y de frecuencia de las tensiones de entrada son variables en el tiempo, como ocurre en los sistemas de generación de energía eléctrica a velocidad variable, que utilizan la energía del viento como fuente primaria de energía. El sistema de control para el trabajo experimental, además de la tarjeta del convertidor matricial, está compuesto por una tarjeta de arreglos de compuertas programable (FPGA y por un procesador digital de señales (DSP con una tarjeta de interfaz gráfica. Abstract: This paper presents the direct and easy way to implement digitally a Sünter-Clare modulation algorithm for a 7.5 KVA, three-wires, matrix converter. This modulation algorithm is used to calculate 3x3 matrix converter duty cycles to produce constant output voltage and frequency signals. The Sünter-Clare modulation algorithm recalculates switching patterns and switching times every sampling period in order to compensate the input voltage and frequency variations. This algorithm is defined in terms of the three-phase input and the output reference voltages at each sampling instant and is convenient for closed loop operations when the input voltage and frequency are variable in

Analysis of multi-dimensional and countercurrent effects in a BWR loss-of-coolant accident

International Nuclear Information System (INIS)

Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

1989-01-01

The presence of parallel enclosed channels in a BWR provides opportunities for multiple flow regimes in co-current and countercurrent flow under Loss-of-Coolant Accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the Steam Sector Test Facility (SSTF) which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests and integral transients with vessel blowdown and refill were performed. The present of multi-dimensional and parallel channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

Algoritmo de cifrado simétrico AES : Aceleración de tiempo de cómputo sobre arquitecturas multicore

OpenAIRE

Pousa, Adrián

2011-01-01

El objetivo de este trabajo es mostrar la aceleración en el tiempo de cómputo del algoritmo criptográfico Advanced Encryption Standard (AES) con clave de tamaño 128bits, que se obtiene al aprovechar el paralelismo que proveen las arquitecturas multicores actuales utilizando herramientas de programación paralela. AES es uno de los algoritmos de criptografía más usados en la actualidad, con el crecimiento de las redes y la información que se maneja hoy en día puede ser necesario cifra...

Datalogger usando nios ii

OpenAIRE

Campoverde Rugel, Luis Enrique; Velásquez Vargas, Washington Adrián; Ponguillo, Ronald

2013-01-01

El presente proyecto consiste en la implementación de un Datalogger utilizando el microprocesador NIOS II el cual fue embebido en el FPGA CYCLONE II que se encuentra integrada en la tarjeta de desarrollo ALTERA DE2, el cual obtiene datos de distintos sensores y los almacena en una tarjeta SD Card. Para la realización del proyecto se aplican cuatro etapas. La primera etapa está basada en obtener los datos mediante el uso de sensores y la transmisión usando un PIC, la siguiente etapa se basa...

Assessment of two BWR accident management strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Petek, M.

1994-01-01

Candidate mitigative strategies for the management of in-vessel events during the late phase (after-core degradation has occurred) of postulated boiling water reactor (BWR) severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities, and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for further assessment. The first was a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertained to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose was to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies were performed during 1991 and this paper provides a discussion of the motivation for and purpose of these strategies, and the potential for their success. ((orig.))

BWR plant dynamic analysis code BWRDYN user's manual

International Nuclear Information System (INIS)

Yokobayashi, Masao; Yoshida, Kazuo; Fujiki, Kazuo

1989-06-01

Computer code BWRDYN has been developed for thermal-hydraulic analysis of a BWR plant. It can analyze the various types of transient caused by not only small but also large disturbances such as operating mode changes and/or system malfunctions. The verification of main analytical models of the BWRDYN code has been performed with measured data of actual BWR plant. Furthermore, the installation of BOP (Balance of Plant) model has made it possible to analyze the effect of BOP on reactor system. This report describes on analytical models and instructions for user of the BWRDYN code. (author)

Modern technology applied in the advanced BWR (ABWR)

International Nuclear Information System (INIS)

Hucik, S.A.

1988-01-01

The advanced boiling water reactor (ABWR) represents the next generation of light water reactors (LWR) to be introduced into commercial operation in the 1990's. The ABWR is the result of the continuing evolution of the BWR, incorporating state-of-the-art technology and improvements based on worldwide experience, and extensive design and test and development programs. This paper discusses how the ABWR development objective focused on an optimized selection of advanced technologies and proven BWR technologies. A technical evaluation of the ABWR shows its superiority in terms of performance characteristics and economics relative to current LWR designs

BWR recirculation pump diagnostic expert system

International Nuclear Information System (INIS)

Chiang, S.C.; Morimoto, C.N.; Torres, M.R.

2004-01-01

At General Electric (GE), an on-line expert system to support maintenance decisions for BWR recirculation pumps for nuclear power plants has been developed. This diagnostic expert system is an interactive on-line system that furnishes diagnostic information concerning BWR recirculation pump operational problems. It effectively provides the recirculation pump diagnostic expertise in the plant control room continuously 24 hours a day. The expert system is interfaced to an on-line monitoring system, which uses existing plant sensors to acquire non-safety related data in real time. The expert system correlates and evaluates process data and vibration data by applying expert rules to determine the condition of a BWR recirculation pump system by applying knowledge based rules. Any diagnosis will be automatically displayed, indicating which pump may have a problem, the category of the problem, and the degree of concern expressed by the validity index and color hierarchy. The rules incorporate the expert knowledge from various technical sources such as plant experience, engineering principles, and published reports. These rules are installed in IF-THEN formats and the resulting truth values are also expressed in fuzzy terms and a certainty factor called a validity index. This GE Recirculation Pump Expert System uses industry-standard software, hardware, and network access to provide flexible interfaces with other possible data acquisition systems. Gensym G2 Real-Time Expert System is used for the expert shell and provides the graphical user interface, knowledge base, and inference engine capabilities. (author)

Economic analysis of hydride fueled BWR

International Nuclear Information System (INIS)

Ganda, F.; Shuffler, C.; Greenspan, E.; Todreas, N.

2009-01-01

The economic implications of designing BWR cores with hydride fuels instead of conventional oxide fuels are analyzed. The economic analysis methodology adopted is based on the lifetime levelized cost of electricity (COE). Bracketing values (1970 and 3010 $/kWe) are used for the overnight construction costs and for the power scaling factors (0.4 and 0.8) that correlate between a change in the capital cost to a change in the power level. It is concluded that a newly constructed BWR reactor could substantially benefit from the use of 10 x 10 hydride fuel bundles instead of 10 x 10 oxide fuel bundles design presently in use. The cost saving would depend on the core pressure drop constraint that can be implemented in newly constructed BWRs - it is between 2% and 3% for a core pressure drop constraint as of the reference BWR, between 9% and 15% for a 50% higher core pressure drop, and between 12% and 21% higher for close to 100% core pressure. The attainable cost reduction was found insensitive to the specific construction cost but strongly dependent on the power scaling factor. The cost advantage of hydride fuelled cores as compared to that of the oxide reference core depends only weakly on the uranium and SWU prices, on the 'per volume base' fabrication cost of hydride fuels, and on the discount rate used. To be economically competitive, the uranium enrichment required for the hydride fuelled core needs to be around 10%.

Algoritmos genéticos

Directory of Open Access Journals (Sweden)

José Jesús Martínez Páez

1998-10-01

Full Text Available Esta técnica se basa en el concepto de evolución a través de selección de los mejores individuos, y de los operadores genéticos de selección, reproducción y mutación. Se trata entonces, de definir un espacio de soluciones para el problema que se quiere solucionar, en una cadena de bits. A esto se le conoce como la codificación del cromosoma, donde cada bit, denominado gen  tiene cierto significado especial. Inicialmente el algoritmo genera al azar muchas de estas cadenas o seres, es decir, una población, que luego confronta can un ambiente, que es el problema solucionar o función que se quiere optimizar. De esta confrontación  o evaluación a que se somete cada ser. Se obtiene información sobre cómo se comporto cada uno. A través de métodos aleatorios, pero con probabilidad de selección proporcional a su comportamiento, es decir, a mejor comportamiento mayor probabilidad, se selecciona una nueva población de seres supuestamente mejores que la generación anterior.

Aplicación de algoritmos heurísticos en la construcción de la población inicial de algoritmos genéticos que resuelven el problema de planeamiento de la expansión de la transmisión

Directory of Open Access Journals (Sweden)

Antonio H. Escobar Z.

2011-01-01

Full Text Available En este artículo se analiza el impacto de seleccionar poblaciones iníciales de buena calidad para ser usadas en algoritmos genéticos, con el propósito de obtener mayor velocidad de convergencia y mejor calidad en las soluciones alcanzadas cuando se resuelve el problema del planeamiento de la expansión a largo plazo de los sistemas de transmisión de energía eléctrica. Los sistemas de prueba que se analizan corresponden a sistemas de alta complejidad, tradicionalmente usados en la literatura especializada. Para generar soluciones iníciales de buena calidad se utilizan algoritmos heurísticos constructivos, particularmente los más utilizados en problemas de planeamiento de la expansión de sistemas de transmisión. Se comparan los resultados obtenidos con los que entregan los algoritmos genéticos que usan poblaciones iniciales aleatorias. Los resultados muestran que una población inicial generada en forma heurística permite obtener soluciones de mejor o igual calidad y con esfuerzos computacionales menores, cuando se resuelven sistemas eléctricos de gran complejidad.

Analysis of a BWR direct cycle forced circulation power plants operation

International Nuclear Information System (INIS)

Andrade, G.G. de.

1973-01-01

First, it is established a general view over the operational problems of the BWR direct cycle forced circulation power plants, and then it is analysed the possibility of the utilization of the energy purged from the turbine as an additional energy for the electrical generation. To simulate the BWR power plant and to obtain the solution of the mathematical model it was developed a computer code named ATOR which shows the feasibility of the proposed method. In this way it is shown the possibility to get a better maneuvering allowance for the BWR power plant whenever it is permitted a convenient use of the vapor extracted from the turbine for the feedwater pre-heaters of the reactor. (author)

BWR Water Chemistry Guidelines: 1993 Revision, Normal and hydrogen water chemistry

International Nuclear Information System (INIS)

Karlberg, G.; Goddard, C.; Fitzpatrick, S.

1994-02-01

The goal of water chemistry control is to extend the operating life of the reactor and rector coolant system, balance-of-plant components, and turbines while simultaneously controlling costs to safeguard the continued economic viability of the nuclear power generation investment. To further this goal an industry committee of chemistry personnel prepared guidelines to identify the benefits, risks, and costs associated with water chemistry in BWRs and to provide a template for an optimized water chemistry program. This document replaces the BWR Normal Water Chemistry Guidelines - 1986 Revision and the BWR Hydrogen Water Chemistry Guidelines -- 1987 Revision. It expands on the previous guidelines documents by covering the economic implications of BWR water chemistry control

ALGORITMOS DE APRENDIZAGEM DE MÁQUINA E VARIÁVEIS DE SENSORIAMENTO REMOTO PARA O MAPEAMENTO DA CAFEICULTURA

Directory of Open Access Journals (Sweden)

Carolina Gusmão Souza

Full Text Available A cafeicultura é uma das principais culturas agrícolas do Brasil e realizar o mapeamento e monitoramento desta cultura é fundamental para conhecer sua distribuição espacial. Porém, mapear estas áreas utilizando imagens de Sensoriamento Remoto não é uma tarefa fácil. Sendo assim, este trabalho foi realizado com o objetivo de comparar o uso de diferentes variáveis e algoritmos de classificação para o mapeamento de áreas cafeeiras. O trabalho foi desenvolvido em três áreas diferentes, que são bastante significativas na produção de café. Foram utilizados 5 algoritmos de aprendizagem de máquinas e 7 combinações de variáveis: espectrais, texturais e geométricas, associadas ao processo de classificação. Um total de 105 classificações foram realizadas, 35 classificações para cada uma das áreas. As classificações que não usaram variáveis espectrais não resultaram em bons índices de acurácia. Nas três áreas, o algoritmo que apresentou as melhores acurácias foi o Support vector machine, com acurácia global de 85,33% em Araguari, 87% em Carmo de Minas e 88,33% em Três Pontas. Os piores resultados foram encontrados com o algoritmo Random Forest em Araguari, com acurácia global de 76,66% e com o Naive Bayes em Carmo de Minas e Três Pontas, com 76% e 82% de acerto. Nas três áreas, variáveis texturais, quando associadas às espectrais, melhoraram a acurácia da classificação. O SVM apresentou o melhor desempenho para as três áreas

Development of the BWR Dry Core Initial and Boundary Conditions for the SNL XR2 Experiments; TOPICAL

International Nuclear Information System (INIS)

Ott, L.J.

1994-01-01

The objectives of the Boiling Water Reactor Experimental Analysis and Model Development for Severe Accidents (BEAMD) Program at the Oak Ridge National Laboratory (ORNL) are: (1) the development of a sound quantitative understanding of boiling water reactor (BWR) core melt progression; this includes control blade and channel box effects, metallic melt relocation and possible blockage formation under severe accident conditions, and (2) provision of BWR melt progression modeling capabilities in SCDAP/RELAP5 (consistent with the BWR experimental data base). This requires the assessment of current modeling of BWR core melt progression against the expanding BWR data base. Emphasis is placed upon data from the BWR tests in the German CORA test facility and from the ex-reactor experiments[Sandia National Laboratories (SNL)] on metallic melt relocation and blockage formation in BWRs, as well as upon in-reactor data from the Annular Core Research Reactor (ACRR) DF-4 BWR test (conducted in 1986 at SNL). The BEAMD Program is a derivative of the BWR Severe Accident Technology Programs at ORNL. The ORNL BWR programs have studied postulated severe accidents in BWRs and have developed a set of models specific to boiling water reactor response under severe accident conditions. These models, in an experiment-specific format, have been successfully applied to both pretest and posttest analyses of the DF-4 experiment, and the BWR severe fuel damage (SFD) experiments performed in the CORA facility at the Kernforschungszentrum Karlsruhe (KfK) in Germany, resulting in excellent agreement between model prediction and experiment. The ORNL BWR models have provided for more precise predictions of the conditions in the BWR experiments than were previously available. This has provided a basis for more accurate interpretation of the phenomena for which the experiments are performed. The experiment-specific models, as used in the ORNL DF-4 and CORA BWR experimental analyses, also provide a basis

Thermohydraulic stability coupled to the neutronic in a BWR; Estabilidad termohidraulica acoplada a la neutronica en un BWR

Energy Technology Data Exchange (ETDEWEB)

Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A. [Comision Federal de Electricidad, Central Nucleoelectrica de Laguna Verde, Carretera Cardel-Nautla Km. 42.5, Mpio. Alto Lucero, Veracruz (Mexico); Castlllo D, R. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: gcm9acpp@cfe.gob.mx

2006-07-01

In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the

Flex concept for US-A BWR extended loss of AC power events

International Nuclear Information System (INIS)

Powers, J.; Aoyagi, Y.; Kataoka, K.; Thomas, S.; Mookhoek, B.

2015-09-01

The US-Advanced Boiling Water Reactor (US-A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (Stp 3 and 4) Combined License Application (Cola) and incorporates numerous design and technology enhancements for improved safety performance. Nuclear Innovation North America (NINA) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The Stp 3 and 4 project has finished the US NRC technical review of the Cola, and the final safety evaluation report (FSER) is scheduled to be issued by the US NRC in 2015. Following the accident at the Fukushima Dai-ichi plant, the US-A BWR was reviewed for Beyond Design Basis Event (BDBE) safety using industry and regulatory guidance for US NRC Order EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation of Beyond Design Basis External Events (BDBEE). By virtue of the design approach, the US-A BWR is capable of providing an indefinite coping period for a station blackout. The use of installed systems with extended coping times is a significant advantage of the US-A BWR compared to most of the plants currently operating in the U.S. In addition, the Stp 3 and 4 design incorporates enhancements consistent with the current US industry Diverse and Flexible Coping Strategies (Flex) initiative. The final technical topic requiring review by the US NRC Advisory Committee on Reactor Safeguards was the Flex Integrated Plan submitted by NINA, and this review was successfully completed. This paper summarizes the progress of the US-A BWR in licensing the Flex Integrated Plan for the project, and describes the technology and features of the US-A BWR design that contribute to safety post-Fukushima. It also provides an informational comparison of the design capabilities of the US-A BWR for extreme external events, and relates these capabilities to re

Flex concept for US-A BWR extended loss of AC power events

Energy Technology Data Exchange (ETDEWEB)

Powers, J. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Aoyagi, Y.; Kataoka, K. [Toshiba Corporation, Kawasaki, Kanagawa (Japan); Thomas, S.; Mookhoek, B., E-mail: jim.powers@toshiba.com [Nuclear Innovation North America, Lake Jackson, Texas (United States)

2015-09-15

The US-Advanced Boiling Water Reactor (US-A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (Stp 3 and 4) Combined License Application (Cola) and incorporates numerous design and technology enhancements for improved safety performance. Nuclear Innovation North America (NINA) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The Stp 3 and 4 project has finished the US NRC technical review of the Cola, and the final safety evaluation report (FSER) is scheduled to be issued by the US NRC in 2015. Following the accident at the Fukushima Dai-ichi plant, the US-A BWR was reviewed for Beyond Design Basis Event (BDBE) safety using industry and regulatory guidance for US NRC Order EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation of Beyond Design Basis External Events (BDBEE). By virtue of the design approach, the US-A BWR is capable of providing an indefinite coping period for a station blackout. The use of installed systems with extended coping times is a significant advantage of the US-A BWR compared to most of the plants currently operating in the U.S. In addition, the Stp 3 and 4 design incorporates enhancements consistent with the current US industry Diverse and Flexible Coping Strategies (Flex) initiative. The final technical topic requiring review by the US NRC Advisory Committee on Reactor Safeguards was the Flex Integrated Plan submitted by NINA, and this review was successfully completed. This paper summarizes the progress of the US-A BWR in licensing the Flex Integrated Plan for the project, and describes the technology and features of the US-A BWR design that contribute to safety post-Fukushima. It also provides an informational comparison of the design capabilities of the US-A BWR for extreme external events, and relates these capabilities to re

PROBLEMA DE ALOCAÇÃO DE CONFIABILIDADE-REDUNDÂNCIA USANDO EVOLUÇÃO DIFERENCIAL AUTO-ADAPTATIVA

OpenAIRE

Spengler, Hellen Cristina; Loch, Gustavo Valentim

2017-01-01

Resumo. Esse trabalho considera a otimização não-linear inteira-mista do problema de alocação de confiabilidade-redundância, determinando simultaneamente a confiabilidade e redundância dos componentes de um sistema em série. Grande número de métodos (exatos e metaheurísticos) já abordaram esse problema, apresentando soluções satisfatórias. Nesse trabalho é utilizado um algoritmo de evolução diferencial auto-adaptativa (SaMDE) junto a um algoritmo de busca local. O exemplo numérico resolvido i...

Maximum thermal loading test of BWR fuel assembly

International Nuclear Information System (INIS)

Nakajima, Yoshitaka; Yoshimura, Kunihiro; Nakamura, Satoshi; Ishizuka, Takao.

1987-01-01

Various proving tests on the reliability of nuclear power plants have been conducted at the Nuclear Power Engineering Test Center and at the Japan Power Plant Engineering and Inspection Corporation. The tests were initiated at the request of the Ministry of International Trade and Industry (MITI). Toshiba undertook one of the proving tests on the reliability of nuclear fuel assembly; the maximum thermal loading test of BWR fuel assembly from the Nuclear Power Engineering Test Center. These tests are part of the proving tests mentioned above, and their purpose is to confirm the reliability of the thermal hydraulic engineering techniques. Toshiba has been engaged for the past nine years in the design, fabrication and testing of the equipment. For the project, a test model fuel assembly was used to measure the critical power of the BWR fuel assembly and the void and fluidity of the coolant. From the test results, it has been confirmed that the heat is transferred safely from the fuel assembly to the coolant in the BWR nuclear power plant. In addition, the propriety and reliability of the thermal hydraulic engineering techniques for the fuel assembly have been proved. (author)

Large bundle BWR test CORA-18: Test results

International Nuclear Information System (INIS)

Hagen, S.; Hofmann, P.; Noack, V.; Sepold, L.; Schanz, G.; Schumacher, G.

1998-04-01

The CORA out-of-pile experiments are part of the international Severe Fuel Damage (SFD) Program. They were performed to provide information on the damage progression of Light Water Reactor (LWR) fuel elements in Loss-of-coolant Accidents in the temperature range 1200 C to 2400 C. CORA-18 was the large BWR bundle test corresponding to the PWR test CORA-7. It should investigate if there exists an influence of the BWR bundle size on the fuel damage behaviour. Therefore, the standard-type BWR CORA bundle with 18 fuel rod simulators was replaced by a large bundle with two additional surrounding rows of 30 rods (48 rods total). Power input and steam flow were increased proportionally to the number of fuel rod simulators to give the same initial heat-up rate of about 1 K/s as in the smaller bundles. Emphasis was put on the initial phase of the damage progression. More information on the chemical composition of initial and intermediate interaction products and their relocation behaviour should be obtained. Therefore, power and steam input were terminated after the onset of the temperature escalation. (orig.) [de

Un algoritmo de clasificación incremental basado en los k vecinos más similares para datos mezclados

Directory of Open Access Journals (Sweden)

Guillermo Sánchez-Díaz

2013-01-01

Full Text Available En este trabajo, se presenta un algoritmo de clasificación incremental basado en los k vecinos más similares, el cual permite trabajar con datos mezclados y funciones de semejanza que no necesariamente son distancias. El algoritmo presentado es adecuado para procesar grandes conjuntos de datos, debido a que sólo almacena en la memoria principal de la computadora los k vecinos más similares procesados hasta el paso t, recorriendo una sola vez el conjunto de datos de entrenamiento. Se presentan resultados obtenidos con diversos conjuntos de datos sintéticos y reales.

Algoritmo genético permutacional para el despliegue y la planificación de sistemas de tiempo real distribuidos

OpenAIRE

Azketa, Ekain; Gutiérrez, J. Javier; Di Natale, Marco; Almeida, Luís; Marcos, Marga

2013-01-01

Resumen: El despliegue y la planificación de tareas y mensajes en sistemas de tiempo real distribuidos son problemas NP-difíciles (NP- hard), por lo que no existen métodos óptimos para solucionarlos en tiempo polinómico. En consecuencia, estos problemas son adecuados para abordarse mediante algoritmos genéricos de búsqueda y optimización. En este artículo se propone un algoritmo genético multiobjetivo basado en una codificación permutacional de las soluciones para abordar el despli...

BWR shutdown analyzer using artificial intelligence (AI) techniques

International Nuclear Information System (INIS)

Cain, D.G.

1986-01-01

A prototype alarm system for detecting abnormal reactor shutdowns based on artificial intelligence technology is described. The system incorporates knowledge about Boiling Water Reactor (BWR) plant design and component behavior, as well as knowledge required to distinguish normal, abnormal, and ATWS accident conditions. The system was developed using a software tool environment for creating knowledge-based applications on a LISP machine. To facilitate prototype implementation and evaluation, a casual simulation of BWR shutdown sequences was developed and interfaced with the alarm system. An intelligent graphics interface for execution and control is described. System performance considerations and general observations relating to artificial intelligence application to nuclear power plant problems are provided

BWR condensate filtration studies

International Nuclear Information System (INIS)

Wilson, J.A.; Pasricha, A.; Rekart, T.E.

1993-09-01

Poor removal of particulate corrosion products (especially iron) from condensate is one of the major problems in BWR systems. The presence of activated corrosion products creates ''hot spots'' and increases piping dose rates. Also, fuel efficiency is reduced and the risk of fuel failure is increased by the deposit of corrosion products on the fuel. Because of these concerns, current EPRI guidelines call for a maximum of 2 ppb of iron in the reactor feedwater with a level of 0.5 ppb being especially desirable. It has become clear that conventional deep bed resins are incapable of meeting these levels. While installation of prefilter systems is an option, it would be more economical for plants with naked deep beds to find an improved bead resin for use in existing systems. BWR condensate filtration technologies are being tested on a condensate side stream at Hope Creek Nuclear Generating Station. After two years of testing, hollow fiber filters (HFF) and fiber matrix filters (FMF), and low crosslink cation resin, all provide acceptable results. The results are presented for pressure drop, filtration efficiency, and water quality measurements. The costs are compared for backwashable non-precoat HFF and FMF. Results are also presented for full deep bed vessel tests of the low crosslink cation resin

Afectos y algoritmos alternativos en las aulas de primaria

OpenAIRE

Gamazo Alonso, Natalia

2017-01-01

El presente Trabajo Fin de Grado (TFG) recoge una sencilla investigación sobre la evolución del dominio afectivo matemático de una muestra de niños que trabajan las matemáticas con el método ABN (Abierto Basado en Números) y de una muestra de alumnos que las trabajan mediante algoritmos tradicionales. Junto con el análisis de la evolución de la afectividad hacia las matemáticas a lo largo de los cursos académicos que conforman la Educación Primaria, se establece también una comparativa ent...

Simulador gráfico de algoritmos matemáticos

OpenAIRE

Alves, Luís; Balsa, Carlos; Pereira, Maria João

2016-01-01

O objetivo principal do trabalho de investigação foi desenvolver uma ferramenta computacional que possa ser usada por professores e alunos no ensino dos Métodos Numéricos. A ferramenta, designada por GraSMA (Graphical Simulator of Mathematical Algorithms), permite visualizar no ecrã a execução dos correspondentes algoritmos. Esta ferramenta integra vários softwares open source e baseia-se na anotação automática de código Octave com funções de inspeção que permitem captar a sucessão de valores...

Characteristics of fluctuating pressure generated in BWR main steam lines

International Nuclear Information System (INIS)

Takahashi, Shiro; Okuyama, Keita; Tamura, Akinori

2009-01-01

The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

The BWR vessel and internals project - 2001 and beyond

International Nuclear Information System (INIS)

Wagoner, V.; Mulford, T.

2001-01-01

The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

The BWR vessel and internals project - 2001 and beyond

Energy Technology Data Exchange (ETDEWEB)

Wagoner, V. [Carolina Power and Light, Progress Energy Building, NC (United States); Mulford, T. [Electric Power Research Institute, Palo Alto, CA (United States)

2001-07-01

The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

Advanced chemistry management system to optimize BWR chemistry control

International Nuclear Information System (INIS)

Maeda, K.; Nagasawa, K.

2002-01-01

BWR plant chemistry control has close relationships among nuclear safety, component reliability, radiation field management and fuel integrity. Advanced technology is required to improve chemistry control [1,3,6,7,10,11]. Toshiba has developed TACMAN (Toshiba Advanced Chemistry Management system) to support BWR chemistry control. The TACMAN has been developed as response to utilities' years of requirements to keep plant operation safety, reliability and cost benefit. The advanced technology built into the TACMAN allows utilities to make efficient chemistry control and to keep cost benefit. TACMAN is currently being used in response to the needs for tools those plant chemists and engineers could use to optimize and identify plant chemistry conditions continuously. If an incipient condition or anomaly is detected at early stage, root causes evaluation and immediate countermeasures can be provided. Especially, the expert system brings numerous and competitive advantages not only to improve plant chemistry reliability but also to standardize and systematize know-how, empirical knowledge and technologies in BWR chemistry This paper shows detail functions of TACMAN and practical results to evaluate actual plant. (authors)

Crud deposition modeling on BWR fuel rods

International Nuclear Information System (INIS)

Kucuk, Aylin; Cheng, Bo; Potts, Gerald A.; Shiralkar, Bharat; Morgan, Dave; Epperson, Kenny; Gose, Garry

2014-01-01

Deposition of boiling water reactor (BWR) system corrosion products (crud) on operating fuel rods has resulted in performance-limiting conditions in a number of plants. The operational impact of performance-limiting conditions involving crud deposition can be detrimental to a BWR operator, resulting in unplanned or increased frequency of fuel inspections, fuel failure and associated radiological consequences, operational restrictions including core power derate and/or forced shutdowns to remove failed fuel, premature discharge of individual bundles or entire reloads, and/or undesirable core design restrictions. To facilitate improved management of crud-related fuel performance risks, EPRI has developed the CORAL (Crud DepOsition Risk Assessment ModeL) tool. This paper presents a summary of the CORAL elements and benchmarking results. Applications of CORAL as a tool for fuel performance risk assessment are also discussed. (author)

Redes de Petri: Modelado e implementación de algoritmos para autómatas programables

Directory of Open Access Journals (Sweden)

Luis Diego Murillo

2008-04-01

Full Text Available El presente trabajo es una monografía orientada hacia la utilización del formalismo de las Redes de Petri, propuesto por Carl Petri en la descripción de Sistemas Dinámicos de Eventos Discretos (DEDS. Las Redes de Petri, cuyo acrónimo en inglés es PN, fueron utilizadas inicialmente para el análisis de algoritmos en la computación paralela o concurrente, pero dada la complejidad de los procesos productivos actuales, las PN son un método alternativo de diseño tanto para el procesoindustrial como para el controlador. En este sentido, este estudio hace una revisión de las referencias bibliográficas donde se indica cómo realizar el modelado y la implementación de algoritmos de control en Controladores Lógicos Programables (PLCs por sus siglas en inglés.

Un Algoritmo Genético Especializado en Planeamiento de Redes de Distribución. Parte II. Detalles del algoritmo y su aplicación; A specialized Genetic Algorithm in Distribution Network planning. Part II. Algorithm details and application

Directory of Open Access Journals (Sweden)

Raúl Nicolás Carvajal Pérez

2011-05-01

Full Text Available En la Parte I de este articulo, se expusieron las características de la planificación de redes y su influencia en la selección de métodos específicos en cada uno de los procedimientos para desarrollar los pasos del algoritmo genético especializado. Aquí se exponen las interioridades del algoritmo y el trabajo experimental para fijar la magnitud de los parámetros generales de las poblaciones y la cantidad de éstas. Se presentan casos resueltos durante el ajuste del sistema de cálculo.  The first section explained the specifics characteristics of network planning and the selected method for genetic operators. In this papers, expose details about the algorithm steps, the mathematical procedures and some applications.

Secondary systems of PWR and BWR

International Nuclear Information System (INIS)

Schindler, N.

1981-01-01

The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

Reconocimiento de Huellas Dactilares Usando Características Locales

Directory of Open Access Journals (Sweden)

Gualberto Aguilar

2008-01-01

Full Text Available El reconocimiento de huellas dactilares es uno de los métodos más populares usados con mayor grado de éxito para la identificación de personas. La huella dactilar tiene características únicas llamadas minucias, las cuales son puntos donde los bordes terminan o se dividen. Los sistemas de identificación que usan patrones biométricos de huella dactilar se denominan AFIS (Sistema de Identificación Automático de Huella Dactilar. En este trabajo se realizó un sistema para reconocimiento de huella dactilar usando combinación de Transformada Rápida de Fourier (FFT con Filtros de Gabor para aclarar la imagen y después un novedoso método para el reconocimiento usando características locales.

Development and recent trend of design of BWR nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Kani, J [Tokyo Shibaura Electric Co. Ltd., Kawasaki, Kanagawa (Japan)

1977-11-01

Many improvements have been carried out in BWR nuclear power plants from BWR-1, represented by Dresden No. 1 plant, to the present BWR-6 as the capacity has increased. In Japan, the plants up to BWR-5 have been constructed. In addition, further fine design improvements are being performed in the complete domestic manufacturing of BRWs based on the operational experiences to date. A variety of investigations on the standardization of nuclear power facilities have been progressing under the leadership of Japanese Ministry of International Trade and Industry since 1975. In this standardization, it is intended to forward the plant design taking eight concrete items into consideration, mainly aiming at carrying cut unerringly the maintenance and inspection, reduction of exposure of employees to radiation, and improvements of the rate of operation of plants and equipment reliability. The containment vessel has been developed in three forms, from Mark 1 through 3, adopting the pressure control system consistently since BWR-2. Mark 1 and 2 were constructed in Japan. However, these designs sacrificed the workability and increased radiation exposure during maintenance as a result of placing emphasis on the safety facilities, therefore Toshiba Electric has investigated the advanced Mark 1 type. Its features are the design for improving the work efficiency in a containment vessel, reducing the radiation exposure of workers, shortening plant construction period, and considering the aseismatic capability. In addition, the following themes are being planned as future standardization: (1) electrically driven control rod driving system, (2) improved design of reactor core, and (3) internal pump system as compared with external re-circulation.

Development and recent trend of disign of BWR nuclear power plants

International Nuclear Information System (INIS)

Kani, Jiro

1977-01-01

Many improvements have been carried out in BWR nuclear power plants from BWR-1, represented by Dresden No. 1 plant, to the present BWR-6 as the capacity has increased. In Japan, the plants up to BWR-5 have been constructed. In addition, further fine design improvements are being performed in the complete domestic manufacturing of BRWs based on the operational experiences to date. A variety of investigations on the standardization of nuclear power facilities have been progressing under the leadership of Japanese Ministry of International Trade and Industry since 1975. In this standardization, it is intended to forward the plant design taking eight concrete items into consideration, mainly aiming at carrying cut unerringly the maintenance and inspection, reduction of exposure of employees to radiation, and improvements of the rate of operation of plants and equipment reliability. The containment vessel has been developed in three forms, from Mark 1 through 3, adopting the pressure control system consistently since BWR-2. Mark 1 and 2 were constructed in Japan. However, these designs sacrificed the workability and increased radiation exposure during maintenance as a result of placing emphasis on the safety facilities, therefore Toshiba Electric has investigated the advanced Mark 1 type. Its features are the design for improving the work efficiency in a containment vessel, reducing the radiation exposure of workers, shortening plant construction period, and considering the aseismatic capability. In addition, the following themes are being planned as future standardization: (1) electrically driven control rod driving system, (2) improved design of reactor core, and (3) internal pump system as compared with external re-circulation. (Wakatsuki, Y.)

Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program; Simulacion en 3 dimensiones de un ciclo de 18 meses para un reactor BWR usando el programa Nod3D

Energy Technology Data Exchange (ETDEWEB)

Hernandez, N.; Alonso, G. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)]. E-mail: nhm@nuclear.inin.mx; Valle, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

2004-07-01

The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

Prueba e implementación de algoritmos de control de calidad de datos de temperatura superficial del aire en un contexto operativo

Directory of Open Access Journals (Sweden)

José Araya

2008-04-01

Full Text Available Se presenta una metodología para el cálculo de rangos de temperatura, así como algoritmos de programación simple para la detección de errores obvios en datos meteorológicos con el fin de mostrar cómo un sistema de control de calidad sencillo, en tiempo real, puede implementarse de forma exitosa. Estos algoritmos fueron probados a través de su programación en un grabador de datos, el cual es el núcleo procesador en una estación meteorológica: primero, bajo condiciones controladas; y, luego, en dos estaciones meteorológicas automáticas con capacidad de transmisión en tiempo real. La investigación realizada muestra que estos algoritmos son efectivospara la detección de valores atípicos que de otra manera podrían ser detectados tardíamente y pasar inadvertidos.

Identification and assessment of BWR in-vessel severe accident mitigation strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Kress, T.S.; Cleveland, J.C.; Petek, M.

1992-01-01

This paper briefly describes the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to evaluate the effectiveness and feasibility of current and proposed strategies for BWR severe accident management. These results are described in detail in the just-released report Identification and Assessment of BWR In-Vessel Severe Accident Mitigation Strategies, NUREG/CR-5869, which comprises three categories of findings. First, an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences is combined with a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, two of the four candidate strategies identified by this effort are assessed in detail. These are (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

BWR level estimation using Kalman Filtering approach

International Nuclear Information System (INIS)

Garner, G.; Divakaruni, S.M.; Meyer, J.E.

1986-01-01

Work is in progress on development of a system for Boiling Water Reactor (BWR) vessel level validation and failure detection. The levels validated include the liquid level both inside and outside the core shroud. This work is a major part of a larger effort to develop a complete system for BWR signal validation. The demonstration plant is the Oyster Creek BWR. Liquid level inside the core shroud is not directly measured during full power operation. This level must be validated using measurements of other quantities and analytic models. Given the available sensors, analytic models for level that are based on mass and energy balances can contain open integrators. When such a model is driven by noisy measurements, the model predicted level will deviate from the true level over time. To validate the level properly and to avoid false alarms, the open integrator must be stabilized. In addition, plant parameters will change slowly with time. The respective model must either account for these plant changes or be insensitive to them to avoid false alarms and maintain sensitivity to true failures of level instrumentation. Problems are addressed here by combining the extended Kalman Filter and Parity Space Decision/Estimator. The open integrator is stabilized by integrating from the validated estimate at the beginning of each sampling interval, rather than from the model predicted value. The model is adapted to slow plant/sensor changes by updating model parameters on-line

Investigation of BWR stability in Forsmark 2

International Nuclear Information System (INIS)

Oguma, R.; Reisch, F.; Bergdahl, B.G.; Lorenzen, J.; Aakerhielm, F.; Kellner, S.

1988-01-01

A series of noise measurements have been conducted at the Forsmark-2 reactor during its start-up operation after the revision in 1987. The main purpose was to investigate the BWR stability problem based on noise analysis, i.e. the problem of resonant power oscillation with frequency of about 0.5 Hz, which tends to arise at high power and low core flow condition. The noise analysis was performed to estimate the noise source which gives rise to the power oscillation, to evaluate the stability condition of the Forsmark-2 reactor in terms of the decay ratio (DR), as well as to investigate a safety related problem in connection with the BWR stability. The results indicate that the power oscillation is due to dynamic coupling between the neutron kinetics and thermal-hydraulics via void reactivity feedback. The DR reached as high as ≅ 0.7 at 63% of the rated power and 4100 kg/s of the total core flow. An investigation was made for the noise recording which represents a strong pressure oscillation with a peak frequency at 0.33 Hz. The result suggests that such pressure oscillation, if the peak frequency coincided with that of the resonant power oscillation, might become a cause of scram. The present noise analysis indicates the importance of a BWR on-line surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

Energy Technology Data Exchange (ETDEWEB)

Arai, Kenji; Ebata, Shigeo [Toshiba Corp., Yokohama (Japan)

1997-07-01

This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding of the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.

Reconocimiento de Gestos usando Cámaras de Profundidad

OpenAIRE

JHONSON GARCÍA, FIORELLA ANNETH

2017-01-01

El proyecto pretende detectar un repertorio de gestos a partir de nubes de puntos 3D obtenidas con cámaras de profundidad tipo Kinect. Jhonson García, FA. (2017). Reconocimiento de Gestos usando Cámaras de Profundidad. http://hdl.handle.net/10251/91700 TFGM

Corrosion issues in the BWR and their mitigation for plant life extension

International Nuclear Information System (INIS)

Gordon, B.M.

1988-01-01

Corrosion is a major service life limiting mechanism for both pressurized water reactors (PWRs) and boiling water reactors (BWRs). For the BWR, stress corrosion cracking of piping has been the major source of concern where extensive research has led to a number of qualified remedies and currently > 90% of susceptible welds have been mitigated or replaced. Stress corrosion cracking of reactor internals due to the interaction of irradiation, as discussed elsewhere in this conference, is also a possible life limiting phenomenon. This paper focusses on two corrosion phenomena in the BWR which have only recently been identified as impacting the universal goal of BWR life extension: the general corrosion of containment structures and the erosion-corrosion of carbon steel piping

BWR Assembly Optimization for Minor Actinide Recycling

Energy Technology Data Exchange (ETDEWEB)

G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

2010-03-22

The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

LAPUR5 BWR stability analysis in Kuosheng nuclear power plant

International Nuclear Information System (INIS)

Kunlung Wu; Chunkuan Shih; Wang, J.R.; Kao, L.S.

2005-01-01

Full text of publication follows: Unstable oscillation of a nuclear power reactor core is one of the main reasons that causes minor core damage. Stability analysis needs to be performed to predict the potential problem as early as possible and to prevent core instability events from happening. Nuclear Regulatory Commission (NRC) requests all BWR licensees to examine each core reload and to impose operating limitations, as appropriate, to ensure compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed. Therefore, the core instability is directly related to the fuel design limits. The core and channel DR (decay ratio) calculation are commonly performed to determine system's stability when new fuel designs are introduced in the core. In order to establish the independent analysis technology for BWR licensees and verifications, the Institute of Nuclear Energy Research (INER) has obtained agreement from NRC and implemented the 'Methodology and Procedure for Calculation of Core and Channel Decay Ratios with LAPUR', which was developed by the IBERINCO in 2001. LAPUR5 uses a multi-nodal description of the neutron dynamics, together with a distributed parameter model of the core thermal hydrodynamics to produce a space-dependent representation of the dynamics of a BWR in the frequency domain for small perturbations around a steady state condition. From the output of LAPUR5, the following results are obtained: global core decay ratio, out-of phase core decay ratio, and channel decay ratio. They are key parameters in the determination of BWR core stability

3D simulation of a core operation cycle of a BWR using Serpent; Simulacion 3D de un ciclo de operacion del nucleo de un BWR usando SERPENT

Energy Technology Data Exchange (ETDEWEB)

Barrera Ch, M. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, 07738 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: rionchez@icloud.com [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2016-09-15

This work had the main goal to develop a methodology to obtain the length of an operating cycle of the core of a BWR under different operating states using the Serpent code. The reactor core modeled in Serpent is composed of 444 fuel assemblies (120 with fresh fuels and 324 fuels from previous cycles), 109 cruciform control rods and light water as moderator and coolant. Once the core of the reactor was modeled in Serpent (Three-dimensional) without considering the cruciform control rods, a simulation was carried out with different steps of burning in the operational state with the average values of the fuel temperature (900 K), moderator temperature (600 K) and voids fraction equal to 0.4. In addition, the thermal power considered was 2017 MWt. This operational state was chosen because a previous analysis (not shown in this work) was carried out in 4 types of control cells. The first and second control cell has all of its natural uranium fuel pellets, with control rod and without control rod respectively. The third and fourth control cell types have various types of enrichment, both natural uranium and gadolinium in their fuel pellets, with control rod and without control rod. The conclusion of this previous analysis was that the behavior of the effective multiplication factor along the fuel burnout within the four control cell types was almost unaffected by the fuel temperature but was affected by the voids fraction. Thus, for this operating cycle in the operating state defined above, its length was 14,63052 GW t/Tm. In addition, at the end of this cycle, the decay heat obtained was equal to 116.71 MWt and the inventory of the most important isotopes to be considered was obtained, such as some isotopes of uranium, neptune, plutonium, americium and curio. (Author)

Advances in BWR water chemistry

International Nuclear Information System (INIS)

Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Mary L.

2012-09-01

This paper reviews recent advances in Boiling Water Reactor (BWR) water chemistry control with examples of plant experiences at U.S. designed BWRs. Water chemistry advances provide some of the most effective methods for mitigating materials degradation, reducing fuel performance concerns and lowering radiation fields. Mitigation of stress corrosion cracking (SCC) of materials remains a high priority and improved techniques that have been demonstrated in BWRs will be reviewed, specifically hydrogen injection combined with noble metal chemical addition (NMCA) and the newer on-line noble metal application process (OLNC). Hydrogen injection performance, an important part of SCC mitigation, will also be reviewed for the BWR fleet, highlighting system improvements that have enabled earlier injection of hydrogen including the potential for hydrogen injection during plant startup. Water chemistry has been significantly improved by the application of pre-filtration and optimized use of ion exchange resins in the CP (condensate polishing) and reactor water cleanup (RWCU) systems. EPRI has monitored and supported water treatment improvements to meet water chemistry goals as outlined in the EPRI BWR Water Chemistry Guidelines, particularly those for SCC mitigation of reactor internals and piping, minimization of fuel risk due to corrosion and crud deposits and chemistry control for radiation field reduction. In recent years, a significant reduction has occurred in feedwater corrosion product input, particularly iron. A large percentage of plants are now reporting <0.1 ppb feedwater iron. The impacts to plant operation and chemistry of lower feedwater iron will be explored. Depleted zinc addition is widely practiced across the fleet and the enhanced focus on radiation reduction continues to emphasize the importance of controlling radiation source term. In addition, shutdown chemistry control is necessary to avoid excessive release of activated corrosion products from fuel

Algoritmo Novedoso Para la Detección de Tareas Repetitivas en el Teclado

Directory of Open Access Journals (Sweden)

Bairon Londo

2015-07-01

Full Text Available En este artículo una herramienta para la detección de tareas repetitivas con secuencias lógicas realizadas a través de comandos del teclado, mediante el diseño e implementación de un algoritmo basado en el uso de autómatas finitos determinísticos y agentes de búsqueda de patrones. La novedad del algoritmo desarrollado radica en que está orientado a la detección de tareas repetitivas cuyas actividades tienen una secuencia lógica y que actualmente no se encuentran automatizadas por lo complejo que es esta labor   Abstract This paper propose a tool  for the detection of repetitive tasks with logical sequences realized across command of the keyboard is proposed, by means of the design and implementation of an algorithm based on the use of finite automata deterministic and agents of bosses' search. The innovation of the developed algorithm takes root in that it is orientated to the detection of repetitive tasks which activities have a logical sequence and that nowadays are not automated by the complex thing that is this labor.

Seismic proving test of BWR primary loop recirculation system

International Nuclear Information System (INIS)

Sato, H.; Shigeta, M.; Karasawa, Y.

1987-01-01

The seismic proving test of BWR Primary Loop Recirculation system is the second test to use the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory. The purpose of this test is to prove the seismic reliability of the primary loop recirculation system (PLR), one of the most important safety components in the BWR nuclear plants, and also to confirm the adequacy of seismic analysis method used in the current seismic design. To achieve the purpose, the test was conducted under conditions and scale as near as possible to actual systems. The strength proving test was carried out with the test model mounted on the vibration table in consideration of basic design earthquake ground motions and other conditions to confirm the soundness of structure and the strength against earthquakes. Detailed analysis and analytic evaluation of the data obtained from the test was conducted to confirm the adequacy of the seismic analysis method and earthquake response analysis method used in the current seismic design. Then, on the basis of the results obtained, the seismic safety and reliability of BWR primary loop recirculation of the actual plants was fully evaluated

Complete BWR--EM LOCA analysis using the WRAP--EM system

International Nuclear Information System (INIS)

Beckmeyer, R.R.; Gregory, M.V.; Buckner, M.R.

1979-01-01

The Water Reactor Analysis Package, Evaluation Model (WRAP--EM), provides a complete analysis of postulated loss-of-coolant accidents (LOCA's) in light--water nuclear power reactors. The system is being developed at the Savannah River Laboratory (SRL) for use by the Nuclear Regulatory Commission (NRC) to interpret and evaluate reactor vendor, evaluation model (EM) analyses. The initial version of the WRAP--EM system for analysis of boiling water reactors (BWR's) is operational. To demonstrate the complete capability of the WRAP--BWR--EM system, a LOCA analysis has been performed for the Hope Creek Plant

Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly

International Nuclear Information System (INIS)

Andrey Ioilev; Maskhud Samigulin; Vasily Ustinenko; Simon Lo; Adrian Tentner

2005-01-01

Full text of publication follows: The goal of this project is to develop an advanced Computational Fluid Dynamics (CFD) computer code (CFD-BWR) that allows the detailed analysis of the two-phase flow and heat transfer phenomena in a Boiling Water Reactor (BWR) fuel bundle under various operating conditions. This code will include more fundamental physical models than the current generation of sub-channel codes and advanced numerical algorithms for improved computational accuracy, robustness, and speed. It is highly desirable to understand the detailed two-phase flow phenomena inside a BWR fuel bundle. These phenomena include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for the analysis of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is still too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Recent progress in Computational Fluid Dynamics (CFD), coupled with the rapidly increasing computational power of massively parallel computers, shows promising potential for the fine-mesh, detailed simulation of fuel assembly two-phase flow phenomena. However, the phenomenological models available in the commercial CFD programs are not as advanced as those currently being used in the sub-channel codes used in the nuclear industry. In particular, there are no models currently available which are able to reliably predict the nature of the flow regimes, and use the appropriate sub-models for those flow regimes. The CFD-BWR code is being developed as a customized module built on the foundation of the commercial CFD Code STAR-CD which provides general two-phase flow modeling capabilities. The paper describes the model development strategy which has been adopted by the development team for the

Limerick BWR turbine control and protection system upgrade success

International Nuclear Information System (INIS)

Tang, C.K.; Pietryka, T.S.; Federico, P.A.; Williams, J.C.

2015-01-01

Westinghouse and Exelon have successfully implemented a digital electro-hydraulic control (DEHC) at Limerick BWR Unit 1 Station to perform the turbine control, protection and reactor pressure functions. The DEHC replaces analog controls and addressed system performance, obsolescence and reliability. This was a first-of-a-kind application for control and protection of the main turbine and BWR pressure control for the distributed control system utilized. The demolition of analog equipment, main control room and front standard modifications, and acceptance testing were completed on schedule during the normal 2014 outage. Key aspects of the project that facilitated this success will be discussed and presented. (author)

Limerick BWR turbine control and protection system upgrade success

Energy Technology Data Exchange (ETDEWEB)

Tang, C.K.; Pietryka, T.S.; Federico, P.A., E-mail: tangck@westinghouse.com, E-mail: pietryt@westinghouse, E-mail: federipa@westinghouse.com [Westinghouse Electric Company, LLC, Cranberry Township, PA (United States); Williams, J.C., E-mail: Jonathan.Williams@exeloncorp.com [Exelon Nuclear, Warrenville, IL (United States)

2015-07-01

Westinghouse and Exelon have successfully implemented a digital electro-hydraulic control (DEHC) at Limerick BWR Unit 1 Station to perform the turbine control, protection and reactor pressure functions. The DEHC replaces analog controls and addressed system performance, obsolescence and reliability. This was a first-of-a-kind application for control and protection of the main turbine and BWR pressure control for the distributed control system utilized. The demolition of analog equipment, main control room and front standard modifications, and acceptance testing were completed on schedule during the normal 2014 outage. Key aspects of the project that facilitated this success will be discussed and presented. (author)

METODOLOGÍA BASADA EN LOS ALGORITMOS VEGA Y MOGA PARA SOLUCIONAR UN PROBLEMA MULTIOBJETIVO EN UN SISTEMA DE PRODUCCIÓN JOB SHOP

Directory of Open Access Journals (Sweden)

Germán Augusto Coca Ortegón

Full Text Available En este artículo se presenta una metodología que pretende minimizar de forma simultánea, en un ambiente de producción tipo "job shop" correspondiente a una empresa metalmecánica, las siguientes variables: tiempo de proceso, costo de mano de obra directa y, asimismo la fracción defectuosa generada por la fatiga del operario. Con este propósito se fusionan elementos de los algoritmos genéticos Vega y Moga, desarrollando para el efecto las siguientes etapas: generar la población inicial, conformar la nueva población, realizar análisis de varianza y por último, comparar con un método híbrido entre sumas ponderadas y algoritmos genéticos. De acuerdo con lo anterior, al evaluar el individuo de menor tiempo de proceso proveniente de la metodología basada en los algoritmos Vega y Moga, respecto al individuo de menor tiempo de desarrollo proveniente del método híbrido entre sumas ponderadas y algoritmos genéticos, se encuentra que el primero supera en desempeño al segundo así: en cuanto a la variable tiempo de proceso (en horas en 27,86%; en cuanto a la variable tiempo de proceso (en semanas en 1,25%; en cuanto a la variable costo de mano de obra directa (MOD en 6,73% y, en cuanto a la variable fracción defectuosa en 25,85%.

Tritium in liquid phase in a BWR-5 like Laguna Verde

International Nuclear Information System (INIS)

Tijerina S, F.; Vargas A, A.; Cardenas J, J.

2011-11-01

In boiling water reactors (BWR), the tritium (H 3 ) takes place mainly as a result of ternary fissions in the nuclear reactors, of those which 75% are in gaseous form and 25% in liquid form. In the liquid phase, the tritium is transported to the pipes of the primary coolant toward condensed tanks or tanks of drainage excesses of radioactive equipment, located in external areas of a BWR, as well as to the processes of radioactive wastes to be able to be directed to the liquid effluents. For that reason, is necessary to know the possible routes of the transport and processes of the tritium in a BWR to control this radioisotope in the site of the event of leaks in equipment s and buried pipes, avoiding that emigrates toward underground flowing and an impact to the environment and to the people in general. (Author)

Behavior of small-sized BWR fuel under reactivity initiated accident conditions

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Fujishiro, Toshio; Horiki, Oichiro; Chen Dianshan; Takeuchi, Kiyoshi.

1992-01-01

The present work was performed on this small-sized BWR fuel, where Zr liner and rod prepressurization were taken as experimental parameters. Experiment was done under simulated reactivity initiated accident (RIA) conditions at Nuclear Safety Research Reactor (NSRR) belonged to Japan Atomic Energy Research Institute (JAERI). Major remarks obtained are as follows: (1) Three different types of the fuel rods consisted of (a) Zr lined/pressurized (0.65MPa), (b) Zr lined/non-pressurized and (c) non-Zr lined/pressurized (o.65MPa) were used, respectively. Failure thresholds of these were not less than that (260 cal/g·fuel) described in Japanese RIA Licensing Guideline. Small-sized BWR and conventional 8 x 8 BWR fuels were considered to be in almost the same level in failure threshold. Failure modes of the three were (a) cladding melt/brittle, (b) cladding melt/brittle and (c) rupture by large ballooning, respectively. (2) The magnitude of pressure pulse at fuel fragmentation was also studied by lined/pressurized and non-lined/pressurized fuels. Above the energy deposition of 370 cal/g·fuel, mechanical energy (or pressure) was found to be released from these fragmented fuels. No measurable difference was, however, observed between the tested fuels and NSRR standard (and conventional 8 x 8 BWR) fuels. (3) It is worthy of mentioning that Zr liner tended to prevent the cladding from large ballooning. Non-lined/pressurized fuel tended to cause wrinkle deformation at cladding. Hence, cladding external was notched much by the wrinkles. (4) Time to fuel failure measured from the tested BWR fuels (pressurization < 0.6MPA) was longer than that measured from PWR fuels (pressurization < 3.2MPa). The magnitude of the former was of the order of 3 ∼ 6s, while that of the latter was < 1s. (J.P.N.)

BWR stability: history and state-of-the-art

International Nuclear Information System (INIS)

Yadigaroglu, George

2014-01-01

The paper briefly recalls the historical developments, reviews the important phenomena, the analytical and simulation tools that are used for the analysis of BWR stability focussing on the linear, frequency domain methods

Class Schedule Assignment Based on Students Learning Rhythms Using A Genetic Algorithm Asignación de horarios de clase basado en los ritmos de aprendizaje de los estudiantes usando un algoritmo genético

Directory of Open Access Journals (Sweden)

Victor F. Suarez Chilma

2013-03-01

Full Text Available The objective of this proposal is to implement a school day agenda focused on the learning rhythms of students of elementary and secondary schools using a genetic algorithm. The methodology of this proposal takes into account legal requirements and constraints on the assignment of teachers and classrooms in public educational institutions in Colombia. In addition, this proposal provides a set of constraints focused on cognitive rhythms and subjects are scheduled at the most convenient times according to the area of knowledge. The genetic algorithm evolves through a process of mutation and selection and builds a total solution based on the best solutions for each group. Sixteen groups in a school are tested and the results of class schedule assignments are presented. The quality of the solution obtained through the established approach is validated by comparing the results to the solutions obtained using another algorithm.El objetivo de esta propuesta es implementar un horario escolar que tenga en cuenta los ritmos de aprendizaje en los estudiantes de educación primaria y secundaria, utilizando un algoritmo genético. La metodología considera los requerimientos legales y las restricciones necesarias para la asignación de maestros y aulas en instituciones educativas públicas de Colombia. Adicionalmente, se establecen un conjunto de restricciones relacionadas con el enfoque en los ritmos cognitivos, determinando las horas de la jornada en las que es más conveniente la ubicación de ciertas materias de acuerdo al área del conocimiento al que pertenecen. El algoritmo genético evoluciona mediante un proceso de mutación y selección, a través del cual se construye una solución completa a partir de la búsqueda de las mejores soluciones por grupo. Se presentan los resultados de las pruebas realizadas para la asignación de una institución con 16 grupos. La calidad de las soluciones obtenidas de acuerdo al enfoque establecido es validada

Parallel channel effects under BWR LOCA conditions

International Nuclear Information System (INIS)

Suzuki, H.; Hatamiya, S.; Murase, M.

1988-01-01

Due to parallel channel effects, different flow patterns such as liquid down-flow and gas up-flow appear simultaneously in fuel bundles of a BWR core during postulated LOCAs. Applying the parallel channel effects to the fuel bundle, water drain tubes with a restricted bottom end have been developed in order to mitigate counter-current flow limiting and to increase the falling water flow rate at the upper tie plate. The upper tie plate with water drain tubes is an especially effective means of increasing the safety margin of a reactor with narrow gaps between fuel rods and high steam velocity at the upper tie plate. The characteristics of the water drain tubes have been experimentally investigated using a small-scaled steam-water system simulating a BWR core. Then, their effect on the fuel cladding temperature was evaluated using the LOCA analysis program SAFER. (orig.)

Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

Energy Technology Data Exchange (ETDEWEB)

Clayton, Dwight A [ORNL; Poore III, Willis P [ORNL

2015-01-01

The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

LBB application in Swedish BWR design

Energy Technology Data Exchange (ETDEWEB)

Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P. [ABB Atom, Vaesteras (Sweden)

1997-04-01

The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.

LBB application in Swedish BWR design

International Nuclear Information System (INIS)

Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P.

1997-01-01

The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions

Evaluation on transmutation performance of minor actinides with high-flux BWR

International Nuclear Information System (INIS)

Setiawan, M.B.; Kitamoto, A.; Taniguchi, A.

2001-01-01

The performance of high-flux BWR (HFBWR) for burning and/or transmutation (B/T) treatment of minor actinides (MA) and long-lived fission products (LLFP) was discussed herein for estimating an advanced waste disposal with partitioning and transmutation (P and T). The concept of high-flux B/T reactor was based on a current 33 GWt-BWR, to transmute the mass of long-lived transuranium (TRU) to short-lived fission products (SLFP). The nuclide selected for B/T treatment was MA (Np-237, Am-241, and Am-243) included in the discharged fuel of LWR. The performance of B/T treatment of MA was evaluated by a new function, i.e. [F/T ratio], defined by the ratio of the fission rate to the transmutation rate in the core, at an arbitrary burn-up, due to all MA nuclides. According to the results, HFBWR could burn and/or transmute MA nuclides with higher fission rate than BWR, but the fission rate did not increase proportionally to the flux increment, due to the higher rate of neutron adsorption. The higher B/T fraction of MA would result in the higher B/T capacity, and will reduce the units of HFBWR needed for the treatment of a constant mass of MA. In addition, HFBWR had a merit of higher mass transmutation compared to the reference BWR, under the same mass loading of MA

Solution to the transport equation with anisotropic dispersion in a BWR type assembly using the AZTRAN code; Solucion de la ecuacion de transporte con dispersion anisotropica en un ensamble tipo BWR usando el codigo AZTRAN

Energy Technology Data Exchange (ETDEWEB)

Chepe P, M. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Xolocostli M, J. V.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: liaison.web@gmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. San Pedro Zacatenco, 07730 Ciudad de Mexico (Mexico)

2016-09-15

Due to the current computing power, the deterministic codes for analyzing nuclear reactors that have been used for several years are becoming more relevant, since much more precise solution techniques can be used; the last century would have been very difficult, since memory and processor capacities were very limited or had high prices on the components. In this work we analyze the effect of the anisotropic dispersion of the effective dispersion section, compared to the isotropic dispersion. The anisotropy implementation was carried out in the AZTRAN transport code, which is part of the AZTLAN platform for nuclear reactors analysis (in development). The AZTRAN code solves the Boltzmann transport equation in one, two and three dimensions at steady state, using the multi-group technique for energy discretization, the RTN-0 nodal method in spatial discretization and for angular discretization the discrete ordinates without considering anisotropy originally. The effect of the anisotropy dispersion on the effective multiplication factor and the axial and radial power on a fuel assembly BWR type are analyzed. (Author)

Verification of a BWR code package by gamma scan measurements

International Nuclear Information System (INIS)

Nakajima, Tsuyoshi; Iwamoto, Tatsuya; Kumanomido, Hironori

1996-01-01

High-burnup 8 x 8 fuel with a large central water rod (called step 2 fuel) has been recently introduced to the latest Japanese boiling water reactor (BWR) plants. Lanthanum-140 gamma intensity is almost directly related to nodal powers. By gamma scan measurement, the axial distribution of 140 La in the exposed fuel was measured at the end of cycle (EOC) 1 and was compared with the calculation by a BWR code package TGBLA/LOGOS. The multienrichment fuel-type core (MEC) design was adopted for the initial cycle core of the plants. The MEC design contains three different enrichment types of fuels to simulate the equilibrium cycles, achieve much higher discharge exposure, and save fuel cycle cost, and the low-enrichment fuels are loaded in periphery and in control cells. Such MEC design could be a challenge to the BWR design methods because of the large spectrum mismatch among the fuel assemblies of the different enrichments. The aforementioned comparison has shown that the accuracy of the TGBLA/LOGOS code package is satisfactory

Comparative analysis of mechanical characteristics of solidified concentrates from BWR system using Yugoslav and Italian cements

International Nuclear Information System (INIS)

Plecas, I.; Peric, A.; Drljaca, J.; Kostadinovic, A.

1987-01-01

In this paper, properties of Italian and Yugoslav cement mixture with BWR evaporation concentrates were compared, research was held upon fifteen samples, according to the adequate formulations. Samples were made in standard cube form, side 10 cm. Functional relationship between decreasing the compressive strength and amount of incorporated BWR concentrate cement mixture was developed. The results of research showed nearly the same mechanical properties of solidified BWR concentrate with Italian and Yugoslav cements. (author)

Projeto de acopladores ópticos utilizando algoritmo genético

OpenAIRE

Jaqueline Oliveira Zampronio

2016-01-01

Resumo: Este projeto é uma proposta de um novo acoplador óptico para a aplicação em circuitos ópticos integrados. O dispositivo foi desenvolvido utilizando o método dos elementos finitos vetorial 3D associado com o algoritmo genético e leva em consideração guias de onda com diferentes altura, largura e índices de refração. O comprimento total do acoplador óptico é de apenas 4,22 µm e o seu funcionamento é baseado em pequenos segmentos cilíndricos que interferem na propagação da onda eletromag...

Predictive uncertainty reduction in coupled neutron-kinetics/thermal hydraulics modeling of the BWR-TT2 benchmark

Energy Technology Data Exchange (ETDEWEB)

Badea, Aurelian F., E-mail: aurelian.badea@kit.edu [Karlsruhe Institute of Technology, Vincenz-Prießnitz-Str. 3, 76131 Karlsruhe (Germany); Cacuci, Dan G. [Center for Nuclear Science and Energy/Dept. of ME, University of South Carolina, 300 Main Street, Columbia, SC 29208 (United States)

2017-03-15

Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.

Predictive uncertainty reduction in coupled neutron-kinetics/thermal hydraulics modeling of the BWR-TT2 benchmark

International Nuclear Information System (INIS)

Badea, Aurelian F.; Cacuci, Dan G.

2017-01-01

Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.

Combinación entre Algoritmos Genéticos y Aleatorios para la Programación de Horarios de Clases basado en Ritmos Cognitivos

OpenAIRE

Castrillón, Omar D

2014-01-01

Se ha diseñado un método basado en algoritmos evolutivos (genéticos y aleatorios) para programar los horarios de clases en una universidad. Esta metodología considera los ritmos cognitivos de los estudiantes que indican que es mejor enseñar algunas asignaturas en intervalos específicos de tiempo. Primero se describen las diferentes técnicas empleadas para desarrollar este problema. Luego se propone una nueva metodología basada en ritmos cognitivos y algoritmos evolutivos, para resolver todas ...

Algoritmo para monitoramento da incidência da malária na Amazônia brasileira, 2003 a 2010

Directory of Open Access Journals (Sweden)

Rui Moreira Braz

2014-03-01

Full Text Available OBJETIVO: Avaliar um algoritmo para detecção da variação da incidência da malária nos municípios da Amazônia brasileira. MÉTODOS: Avaliou-se um sistema de monitoramento automatizado, baseado em um algoritmo desenvolvido anteriormente pelos autores. O algoritmo utiliza o diagrama de controle por quartis para classificação dos municípios em quatro grupos, conforme a variação da incidência da malária: grupo 1 (redução da incidência-aqueles com incidência abaixo dos valores esperados; grupo 2 (incidência esperada-aqueles com incidência dentro dos valores esperados; grupo 3 (epidemia-aqueles com incidência acima dos valores esperados; grupo 4 (caso esporádico-aqueles com apenas um caso durante o ano. O período de análise foi de 2003 a 2010. Foram estudados todos os municípios existentes nos nove estados que compõem a Amazônia brasileira (805 municípios em 2003 e 807 a partir de 2004. RESULTADOS: Com base nessa metodologia, os municípios da região foram assim classificados: grupo 1, 152 municípios (18,8% da região em 2003 e 109 (13,5% em 2010; grupo 2, 206 (25,6% em 2003 e 331 (41,0% em 2010; grupo 3, 391 (48,6% em 2003 e 308 (38,2% em 2010; e grupo 4, 56 (7,0% em 2003 e 59 (7,3% em 2010. CONCLUSÕES: O algoritmo possibilitou verificar que, na Amazônia brasileira, o número de municípios com epidemias de malária em 2010 diminuiu em relação a 2003, enquanto o número de municípios com incidência esperada aumentou. No mesmo período, houve pouca variação no número de municípios com redução da incidência e daqueles com casos esporádicos.

Specifications of the BWR simulator for HAMMLAB 2000

International Nuclear Information System (INIS)

Grini, Rolf-Einar; Miettinen, Jaakko; Nurmilaukas, Pekka; Raussi; Pekka; Saarni, Ray; Stokke; Egil; Soerensen, Aimar; Tiihonen, Olli

1998-02-01

The Boiling Water Reactor (BWR) simulator for HAMMLAB 2000 will be a model of the Swedish plant Forsmark-3. This report gives the specifications of the BWR simulator. The bulk of the report is a copy of the relevant addendum to the contract with the developer, and to the contract with the group of utilities and with ABB Atom. After a general overview, each plant system is described one after the other (using the reference plant system coding), and the simulation of each system is specified. Even the systems that shall not be simulated are included; in those cases the specification is: It is not required that ... is simulated. A list of malfunctions is given, as well as a list of validation transients. Finally the operator interface is specified. (author)

Standard Technical Specifications, General Electric plants, BWR/4

International Nuclear Information System (INIS)

1992-09-01

This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power specifications. This report contains three volumes. This document, Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS

BWR zinc addition Sourcebook

International Nuclear Information System (INIS)

Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Alfred J.

2014-01-01

Boiling Water Reactors (BWRs) have been injecting zinc into the primary coolant via the reactor feedwater system for over 25 years for the purpose of controlling primary system radiation fields. The BWR zinc injection process has evolved since the initial application at the Hope Creek Nuclear Station in 1986. Key transitions were from the original natural zinc oxide (NZO) to depleted zinc oxide (DZO), and from active zinc injection of a powdered zinc oxide slurry (pumped systems) to passive injection systems (zinc pellet beds). Zinc addition has continued through various chemistry regimes changes, from normal water chemistry (NWC) to hydrogen water chemistry (HWC) and HWC with noble metals (NobleChem™) for mitigation of intergranular stress corrosion cracking (IGSCC) of reactor internals and primary system piping. While past reports published by the Electric Power Research Institute (EPRI) document specific industry experience related to these topics, the Zinc Sourcebook was prepared to consolidate all of the experience gained over the past 25 years. The Zinc Sourcebook will benefit experienced BWR Chemistry, Operations, Radiation Protection and Engineering personnel as well as new people entering the nuclear power industry. While all North American BWRs implement feedwater zinc injection, a number of other BWRs do not inject zinc. This Sourcebook will also be a valuable resource to plants considering the benefits of zinc addition process implementation, and to gain insights on industry experience related to zinc process control and best practices. This paper presents some of the highlights from the Sourcebook. (author)

BWR full integral simulation test (FIST) pretest predictions with TRACBO2

International Nuclear Information System (INIS)

Sutherland, W.A.; Alamgir, M.

1984-01-01

The Full Integral Simulation Test program is a three pronged approach to the development of best-estimate analysis capability for BWR systems. An analytical method development program is underway to extend the BWR-TRAC computer code to model reactor kinetics and major interfacing systems, including balance-of-plant, to improve application modeling flexibility, and to reduce computer running time. An experimental program is underway in a new single bundle system test facility to extend the large break loss-of-coolant accident LOCA data base to small breaks and operational transients. And a method qualification program is underway to test TRACBO2 against experiments in the FIST facility. The recently completed Phase 1 period included a series of LOCA and power transient tests, and successful pretest analysis of the large and small break LOCA tests with TRACBO2. These comparisons demonstrate BWR-TRAC capability for small and large break analysis, and provide detailed understanding of the phenomena

BWR water chemistry impurity studies

International Nuclear Information System (INIS)

Ljungberg, L.G.; Korhonen, S.; Renstroem, K.; Hofling, C.G.; Rebensdorff, B.

1990-03-01

Laboratory studies were made on the effect of water impurities on environmental cracking in simulated BWR water of stainless steel, low alloy steel and nickel-base alloys. Constant elongation rate tensile (CERT) tests were run in simulated normal water chemistry (NWC), hydrogen water chemistry (HWC), or start-up environment. Sulfate, chloride and copper with chloride added to the water at levels of a fraction of a ppM were found to be extremely deleterious to all kinds of materials except Type 316 NG. Other detrimental impurities were fluoride, silica and some organic acids, although acetic acid was beneficial. Nitrate and carbon dioxide were fairly inoccuous. Corrosion fatigue and constant load tests on compact tension specimens were run in simulated normal BWR water chemistry (NWC) or hydrogen water chemistry (HWC), without impurities or with added sulfate or carbon dioxide. For sensitized Type 304 SS in NWC, 0.1 ppM sulfate increased crack propagation rates in constant load tests by up to a factor of 100, and in fatigue tests up to a factor of 10. Also, cracking in Type 316 nuclear grade SS and Alloy 600 was enhanced, but to a smaller degree. Carbon dioxide was less detrimental than sulfate. 3 figs., 4 tabs

Algoritmos Wavenet con Aplicaciones en la Aproximación de Señales: un Estudio Comparativo

Directory of Open Access Journals (Sweden)

C.R. Domínguez Mayorga

2012-10-01

Full Text Available Resumen: En este trabajo de investigación se aplican métodos adaptables en el diseño de algoritmos computacionales, dichos algoritmos emplean redes neuronales y series de wavelets para construir “neuroaproximadores” wavenets. Se muestra cómo las wavenets pueden combinarse con los métodos autosintonizables para obtener el seguimiento de señles complejas que están en función del tiempo. Los algoritmos obtenidos se aplican en la aproximación de señales que representan funciones algebraicas y funciones aleatorias, así como en una señal médica deun ECG. Se muestran los resultados en simulación numérica de dos arquitecturas de neuroaproximadores wavenets: el primero está basado en una wavenet, con el cual se aproximan las señales bajo estudio donde los parámetros de la red neuronal son ajustados en línea; el otro esquema emplea un filtro IIR a la salida de la red wavenet para discriminar las contribuciones de aquellas neuronas que tienen menos peso en la aproximación de la señal, lo que ayuda a reducir el tiempo de convergencia a un error mínimo deseado. Abstract: In this paper adaptable methods for computational algorithms are presented. These algorithms use neural networks and wavelet series to build neuro wavenets approximators. The algorithms obtained are applied to the approximation of signals that represent algebraic functions and random functions, as well as a medical EKG signal. It shows how wavenets can be combined with auto-tuning methods for tracking complex signals that are a function of time. Results are shown in numerical simulation of two architectures of neural approximators wavenets: the first is based on a wavenet with which they approach the signals under study where the parameters of the neural network are adjusted online, the other neuro approximator scheme uses an IIR filter to the output of wavenet, which serves to filter the out- put, in this way

Coretran/Vipre assembly critical power assessment against Nupec BWR experiments

International Nuclear Information System (INIS)

Aounallah, Y.

2001-01-01

This study has been performed, in the framework of the STARS project, to assess CORETRAN-01/VIPRE-02 code capability to predict critical heat flux conditions for BWR fuel assemblies. The assessment is based on comparisons of the code results with the NUPEC steady-state critical power measurements on full-scale assemblies tested under a range of flow conditions. Two assembly types were considered, the standard BWR 8 x 8 and the so-called ''high-burnup'' assembly, similar to GE-10. Code modelling options that have a significant impact on the results have been identified, along with code limitations. (author)

Recent BWR fuel management reactor physics advances

International Nuclear Information System (INIS)

Crowther, R.L.; Congdon, S.P.; Crawford, B.W.; Kang, C.M.; Martin, C.L.; Reese, A.P.; Savoia, P.J.; Specker, S.R.; Welchly, R.

1982-01-01

Improvements in BWR fuel management have been under development to reduce uranium and separative work (SWU) requirements and reduce fuel cycle costs, while also maintaining maximal capacity factors and high fuel reliability. Improved reactor physics methods are playing an increasingly important role in making such advances feasible. The improved design, process computer and analysis methods both increase knowledge of the thermal margins which are available to implement fuel management advance, and improve the capability to reliably and efficiently analyze and design for fuel management advances. Gamma scan measurements of the power distributions of advanced fuel assembly and advanced reactor core designs, and improved in-core instruments also are important contributors to improving 3-d predictive methods and to increasing thermal margins. This paper is an overview of the recent advances in BWR reactor physics fuel management methods, coupled with fuel management and core design advances. The reactor physics measurements which are required to confirm the predictions of performance fo fuel management advances also are summarized

BWR normal water chemistry guidelines: 1986 revision

International Nuclear Information System (INIS)

1988-09-01

Boiling water reactors (BWRs) have experienced stress corrosion cracking in the reactor cooling system piping resulting in adverse impacts on plant availability and personnel radiation exposure. The BWR Owners Group and EPRI have sponsored a major research and development program to provide remedies for this stress corrosion cracking problem. This work shows that the likelihood of cracking depends on the plant's water chemistry performance (environment) as well as on material condition and stress level. Plant experience and other research demonstrate that water quality also affects fuel performance and radiation field buildup in BWRs. This report,''BWR Normal Water Chemistry Guidelines: 1986 Revision,'' presents suggested generic water chemistry specifications, justifies the proposed water chemistry limits, suggests responses to out-of-specification water chemistry, discusses available chemical analysis methods as well as data management and surveillance schemes, and details the management philosophy required to successfully implement a water chemistry control program. An appendix contains recommendations for water quality of auxiliary systems. 73 refs., 20 figs., 9 tabs

Hydrogen injection device in BWR type reactor

International Nuclear Information System (INIS)

Takagi, Jun-ichi; Kubo, Koji.

1988-01-01

Purpose: To reduce the increasing ratio of main steam system dose rate due to N-16 activity due to excess hydrogen injection in the hydrogen injection operation of BWR type reactors. Constitution: There are provided a hydrogen injection mechanism for injecting hydrogen into primary coolants of a BWR type reactor, and a chemical injection device for injecting chemicals such as methanol, which makes nitrogen radioisotopes resulted in the reactor water upon hydrogen injection non-volatile, into the pressure vessel separately from hydrogen. Injected hydrogen and the chemicals are not reacted in the feedwater system, but the reaction proceeds due to the presence of radioactive rays after the injection into the pressure vessel. Then, hydrogen causes re-combination in the downcomer portion to reduce the dissolved oxygen concentration. Meanwhile, about 70 % of the chemicals is supplied by means of a jet pump directly to the reactor core, thereby converting the chemical form of N-16 in the reactor core more oxidative (non-volatile). (Kawakami, Y.)

BWR radiation buildup control with ionic zinc

International Nuclear Information System (INIS)

Marble, W.J.; Wood, C.J.; Leighty, C.E.; Green, T.A.

1986-01-01

In 1983 a hypothesis was disclosed which suggested that the presence of ionic zinc in the reactor water of the BWR could reduce radiation buildup. This hypothesis was developed from correlations of plant data, and subsequently, from laboratory experiments which demonstrated clearly that ionic zinc inhibits the corrosion of stainless steel. The benefits of zinc addition have been measured at the Vallecitos Nuclear Center under and EPRI/GE project. Experimentation and analyses have been performed to evaluate the impact of intentional zinc addition on the IGSCC characteristics of primary system materials and on the performance of the nuclear fuel. It has been concluded that no negative effects are expected. The author conclude that the intentional addition of ionic zinc to the BWR reactor water at a concentration of approximately 10 ppb will provide major benefits in controlling the Co-60 buildup on primary system stainless steel surfaces. The intentional addition of zinc is now a qualified technique for use in BWRs

Algoritmo especializado aplicado ao planejamento da expansÃo de redes aÃreas de sistemas de distribuiÃÃo

OpenAIRE

Willian Douglas Ferrari MendonÃa

2014-01-01

No presente trabalho à apresentado o desenvolvimento de um algoritmo especializado para o planejamento da expansÃo de redes aÃreas de sistemas de distribuiÃÃo. A tÃcnica utilizada para soluÃÃo à a HeurÃstica Construtiva que tem sido utilizada em conjunto com modelos matemÃticos de otimizaÃÃo para resolver o problema. No entanto o algoritmo apresentado nÃo emprega um modelo matemÃtico de otimizaÃÃo, em outras palavras, um modelo composto de funÃÃo objetivo e restriÃÃes. Assim, em vez de trabal...

Diseño e implementación de un algoritmo para translación de protocolo entre las normas IEC60870-5-104 y MODBUS TCP/IP

OpenAIRE

Feria Briceño, John Henry; Universidad Distrital Francisco José de Caldas

2015-01-01

Este artículo muestra los resultados del proceso de diseño, implementación y prueba de un algoritmo para hacer la translación de protocolo entre las normas IEC 60870-5-104 y MODBUS TCP/IP, éste modelo de algoritmos son de amplio uso en los procesos de supervisión y automatización en sub-estaciones eléctricas.

El algoritmo HyRPNI y una aplicación en bioinformática

Directory of Open Access Journals (Sweden)

Gloria Inés Alvarez V.

2011-06-01

Full Text Available Proponemos un algoritmo de inferencia gramatical para lenguajes regulares que permite ahorrar cómputo al usar dos criterios diferentes para elegir los estados a ser procesados, un criterio se usa en la primera fase del proceso de inferencia (al principio y el otro en el resto del proceso. Realizamos experimentos para observar el desempeño del algoritmo, para aprender sobre el tamaño ideal de su primera fase y para mostrar su aplicación en la solución de un problema específico en bioinformática: la predicción de sitios de corte en poliproteínas codificadas por virus de la familia Potyviridae./ We propose a grammar inference algorithm for regular languages which saves computational cost by using two different criteria to choose states to be processed: one in the first phase of the inference process (the beginning and another for the rest of the process. We applied experiments to observe performance of the algorithm, to learn about the best size of its first phase and to show results of its application to solve a specific problem in Bioinformatics: the cleavage site prediction problem in polyproteins encoded by viruses of the Potyviridae family.

Delivering high performance BWR fuel reliably

International Nuclear Information System (INIS)

Schardt, J.F.

1998-01-01

Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

Estudio de Algoritmos 2-Deslizantes Aplicados al Control de Pilas de Combustible

OpenAIRE

Cristian Kunusch; Paul F. Puleston; Miguel A. Mayosky

2008-01-01

Resumen: En este trabajo se hace un estudio comparativo de tres diferentes técnicas de control por modo deslizante de segundo orden, aplicadas al problema específico del control de respiración de una pila de combustible PEM. Los algoritmos diseñados se contrastan por simulación utilizando el modelo completo del sistema, poniendo particular énfasis en la respuesta transitoria y la robustez frente a perturbaciones. Palabras clave: Pilas de Combustible, Control no lineal, Modo Deslizante

Condensate polishing guidelines for PWR and BWR plants

International Nuclear Information System (INIS)

Robbins, P.; Crinigan, P.; Graham, B.; Kohlmann, R.; Crosby, C.; Seager, J.; Bosold, R.; Gillen, J.; Kristensen, J.; McKeen, A.; Jones, V.; Sawochka, S.; Siegwarth, D.; Keeling, D.; Polidoroff, T.; Morgan, D.; Rickertsen, D.; Dyson, A.; Mills, W.; Coleman, L.

1993-03-01

Under EPRI sponsorship, an industry committee, similar in form and operation to other guideline committees, was created to develop Condensate Polishing Guidelines for both PWR and BWR systems. The committee reviewed the available utility and water treatment industry experience on system design and performance and incorporated operational and state-of-the-art information into document. These guidelines help utilities to optimize present condensate polisher designs as well as be a resource for retrofits or new construction. These guidelines present information that has not previously been presented in any consensus industry document. The committee generated guidelines that cover both deep bed and powdered resin systems as an integral part of the chemistry of PWR and BWR plants. The guidelines are separated into sections that deal with the basis for condensate polishing, system design, resin design and application, data management and performance and management responsibilities

Role of BWR MK I secondary containments in severe accident mitigation

International Nuclear Information System (INIS)

Greene, S.R.

1986-01-01

The recent advent of detailed containment analysis codes such as CONTAIN and MELCOR has facilitated the development of the first large-scale, architectural-based BWR secondary containment models. During the past year ORNL has developed detailed, plant-specific models of the Browns Ferry and Peach Bottom secondary containments, and applied these models in a variety of studies designed to evaluate the role and effectiveness of BWR secondary containments in severe accident mitigation. The topology and basis for these models is discussed, together with some of the emerging insights from these studies

Automatic refueling platform and CRD remote handling device for BWR plant

International Nuclear Information System (INIS)

Kato, Hiroaki; Takagi, Kaoru

1978-01-01

In BWR plants, machines for replacing fuel assemblies and control rod drives are usually operated directly by personnel. An automatic refueling platform and a CRD remote handling device aiming at radiation exposure reduction and handling perfectness are described, which are already used in BWR plants. Automation of the former is achieved in transporting fuel assemblies between a reactor pressure vessel and a fuel storage pool, shuffling fuel assemblies in a reactor core and moving fuel assemblies in a fuel storage pool. In the latter, replacement of CRDs is nearly all performed remotely. (Mori, K.)

BWR fuel performance

International Nuclear Information System (INIS)

Baily, W.E.; Armijo, J.S.; Jacobson, J.; Proebstle, R.A.

1979-01-01

The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

Sensitivity of BWR shutdown margin tests to local reactivity anomalies

International Nuclear Information System (INIS)

Cokinos, D.M.; Carew, J.F.

1987-01-01

Successful shutdown margin (SDM) demonstration is a required procedure in the startup of a newly configured boiling water reactor (BWR) core. In its most reactive condition throughout a cycle, a BWR core must be capable of being made subcritical by a specified margin with the highest worth control rod fully withdrawn and all other rods at their fully inserted positions. Two different methods are used to demonstrate SDM: (a) the adjacent-rod test and (b) the in-sequence test. In the adjacent-rod test, the strongest rod is fully withdrawn and an adjacent rod is withdrawn to reach criticality. In the in-sequence test, control rods spread throughout the core are withdrawn in a predetermined sequence of withdrawals. Larger than expected core k/sub eff/ values have been observed during the performance of BWR SDM tests. The purpose of the work summarized in this paper has been to investigated and quantify the sensitivity of both the adjacent-rod and in-sequence SDM tests to local reactivity anomalies. This was accomplished by introducing reactivity perturbations at selected four-bundle cell locations and by evaluating their effect on core reactivity in each of the two tests

Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

International Nuclear Information System (INIS)

Morikawa, Yoshitake

1995-01-01

The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolant system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data

Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Morikawa, Yoshitake [ISOGO Nuclear Engineering Center, Yokohama (Japan)

1995-03-01

The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolant system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data.

Relleno de huecos en mallas 3D completas mediante algoritmos de restauración de imagen

Directory of Open Access Journals (Sweden)

Emiliano Pérez Hernàndez

2012-10-01

Full Text Available Resumen: En este artículo se propone un método para rellenar huecos en objetos 3D. La idea principal es la de utilizar un algoritmo de restauración de imágenes para reconstruir las partes que se han de completar. Para ello es necesario hacer una transformación de los datos de la superficie 3D en datos de imagen 2D. Por tanto, el algoritmo propuesto parte de una etapa inicial de identificación de huecos. A continuación se hace la selección, para cada hueco, de una porción de malla representativa. Seguidamente se calcula el punto de vista o plano de proyección adecuado para obtener una imagen de rango de esa porción de malla. A la imagen de rango obtenida se le aplica el algoritmo de restauración de imagen. Finalmente, se aplica una transformación inversa 2D a 3D y se integra el éresultado con la malla inicial. Se trata de un algoritmo robusto, válido para diversos tipos de huecos y diferentes tamaños de los mismos. Abstract: In this work a method for filling holes in 3D meshes based on a 2D image restoration algorithm is expounded. To do that, data must be converted to a suitable input format, a 3D to 2D transformation is executed by projecting the 3D surface onto a grid. Therefore, the proposed algorithm starts by a first stage of holes identification. Then, a meaningful mesh portiéon is choosen for each hole. Afterward, the suitable plane of projection must be computed to get the range image of the mesh portion. Later, it is applied the restauration image algorithm to the range image. Finally, an inverse transformation 2D to 3D is performed and the new produced data are integrated with the initial mesh. The result is a robust algorithm which works correctly with several kind of holes and for di_erent sizes of them. Palabras clave: visión por computador, gráficos por computador, imágenes digitales, imágenes de rango, restauración de imágenes, Keywords: computer vision, computer graphics, digital images, range images

Study of the Utilization BWR Type Nuclear Power Reactor for Desalination Process

International Nuclear Information System (INIS)

Itjeu Karliana; Sumijanto; Dhandhang Purwadi, M.

2008-01-01

The needs of fresh water increased by rapid population growth and industrials expansion, but these demands can not be prepared naturally. Following this case, seawater desalination becomes the primer option which can fulfill the need through the nuclear desalination technology. The coupled nuclear power reactor enables to supply thermal energy for auxiliary equipment and pumps operation. The utilization study of power reactor type BWR coupled with desalination process has been performed. The goal of study is to obtain characteristic data of desalted water specification which desalination system coupling with nuclear power plant produced energy for desalination process. The study is carried out by browsing data and information, and comprehensive review of thermal energy correlation between NPP with desalination process installation. According to reviewing are found that the thermal energy and electric power utilization from the nuclear power reactor are enable to remove the seawater to produce desalted water and also to operate auxiliary equipments. The assessment results is VK-300 reactor prototype, BWR type 250 MW(e) power are cogeneration unit can supplied hot steam temperature 285 °C to the extraction turbine to empower 150 MW electric power, and a part of hot steam 130 °C is use to operate desalination process and remind heat is distribute to the municipal and offices at that region. The coupled of VK-300 reactor power type BWR with desalination installation of MED type enable to produce desalted water with high quality distillate. Based on the economic calculation that the VK-300 reactor power of BWR type produced water distillate capacity is 300.000 m 3 /hour with cost US$ 0.58/m 3 . The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

Design guideline to prevent the pipe rupture by radiolysis gases in BWR steam piping

International Nuclear Information System (INIS)

Inagaki, T.; Miyagawa, M.; Ota, T.; Sato, T.; Sakata, K.

2009-01-01

In late 2001, pipe rupture accidents due to fast combustion of radiolysis gas occurred in Japan and elsewhere's BWR power plants. TENPES began to set up the guideline as action to such a new problem to prevent accumulation and combustion of radiolysis gas in BWR steam piping. And then, the first edition of guideline was published in October 2005. Afterwards, the experimental study about combustion/detonation of radiolysis gas have been continued. And in March 2007, TENPES published a revised edition of the guideline. This is the report of the revised edition of that guideline. According to this guideline, it became possible to design BWR's steam piping to prevent accumulation of radiolysis gas. (author)

Neutron noise analysis of BWR using time series analysis

International Nuclear Information System (INIS)

Fukunishi, Kohyu

1976-01-01

The main purpose of this paper is to give more quantitative understanding of noise source in neutron flux and to provide a useful tool for the detection and diagnosis of reactor. The space dependent effects of distributed neutron flux signals at the axial direction of two different strings are investigated by the power contribution ratio among neutron fluxes and the incoherent noise spectra of neutron fluxes derived from autoregressive spectra. The signals are measured on the medium sized commercial BWR of 460 MWe in Japan. From the obtained results, local and global noise sources in neutron flux are discussed. This method is indicated to be a useful tool for detection and diagnosis of anomalous phenomena in BWR. (orig./RW) [de

A BWR 24-month cycle analysis using multicycle techniques

International Nuclear Information System (INIS)

Hartley, K.D.

1993-01-01

Boiling water reactor (BWR) fuel cycle design analyses have become increasingly challenging in the past several years. As utilities continue to seek improved capacity factors, reduced power generation costs, and reduced outage costs, longer cycle lengths and fuel design optimization become important considerations. Accurate multicycle analysis techniques are necessary to determine the viability of fuel designs and cycle operating strategies to meet reactor operating requirements, e.g., meet thermal and reactivity margin constraints, while minimizing overall fuel cycle costs. Siemens Power Corporation (SPC), Nuclear Division, has successfully employed multi-cycle analysis techniques with realistic rodded cycle depletions to demonstrate equilibrium fuel cycle performance in 24-month cycles. Analyses have been performed by a BWR/5 reactor, at both rated and uprated power conditions

Coretran/Vipre assembly critical power assessment against Nupec BWR experiments

Energy Technology Data Exchange (ETDEWEB)

Aounallah, Y. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

2001-07-01

This study has been performed, in the framework of the STARS project, to assess CORETRAN-01/VIPRE-02 code capability to predict critical heat flux conditions for BWR fuel assemblies. The assessment is based on comparisons of the code results with the NUPEC steady-state critical power measurements on full-scale assemblies tested under a range of flow conditions. Two assembly types were considered, the standard BWR 8 x 8 and the so-called ''high-burnup'' assembly, similar to GE-10. Code modelling options that have a significant impact on the results have been identified, along with code limitations. (author)

Development of RBWR (Resource-renewable BWR) for environmental burden reduction of radioactive wastes

International Nuclear Information System (INIS)

Hino, Tetsushi; Ohtsuka, Masaya; Moriya, Kumiaki; Matsuura, Masayoshi

2014-01-01

Accumulation of long-life transuranium elements produced as by-products with uranium fuel burning became an issue of nuclear power. Hitachi had been developing the reactor with transuranium elements burning as fuels based on BWR type reactors successfully used as commercial reactors: RBWR (Resource-renewable BWR). Efficient transmutation and fissioning of transuranium elements needed adjustment of in-core neutron energy spectra distribution better for nuclear reaction of transuranium elements. Taking advantage of characteristics of BWR type reactors with neutron spectra hardening more easily adjustable than other type of reactors, multiple recycling and fissioning transuranium elements as fuels could make environmental burden reduction of radioactive wastes and efficient use of resources compatible. This article described the concept and history of RBWR and showed its specifications and reactor core characteristics. (T. Tanaka)

Corrosion resistance improvement of ferritic steels through hydrogen additions to the BWR coolant

International Nuclear Information System (INIS)

Gordon, B.M.; Jewett, C.W.; Pickett, A.E.; Indig, M.E.

1984-01-01

Motivated by the success of oxygen suppression for mitigation of intergranular stress corrosion cracking (IGSCC) in weld sensitized austenitic materials used in Boiling Water Reactors (BWRs), oxygen suppression, through hydrogen additions to the feedwater was investigated to determine its affect on the corrosion resistance of ferritic and martensitic BWR structural materials. The results of these investigations are presented in this paper, where particular emphasis is placed on the corrosion performance of BWR pressure vessel low alloy steels, carbon steel piping materials and martensitic pump materials. It is important to note that the corrosion resistance of these materials in the BWR environment is excellent. Consequently this investigation was also motivated to determine whether there were any detrimental effects of hydrogen additions, as well as to identify any additional margin in ferritic/martensitic materials corrosion performance

Phenomenology of BWR fuel assembly degradation

Science.gov (United States)

Kurata, Masaki; Barrachin, Marc; Haste, Tim; Steinbrueck, Martin

2018-03-01

Severe accidents occurred at the Fukushima-Daiichi Nuclear Power Station (FDNPS) which required an immediate re-examination of fuel degradation phenomenology. The present paper reviews the updated knowledge on the phenomenology of the fuel degradation, focusing mainly on the BWR fuel assembly degradation at the macroscopic scale and that of the individual interactions at the meso-scale. Oxidation of boron carbide (B4C) control rods potentially generates far larger amounts of heat and hydrogen under BWR accident conditions. All integral tests with B4C control rods or control blades have shown early failure, liquefaction, relocation and oxidation of B4C starting at temperatures around 1250 °C, well below the significant interaction temperatures of UO2-Zry. These interactions or reactions potentially influence the progress of fuel degradation in the early phase. The steam-starved conditions, which are being discussed as a likely scenario at the FDNPS accident, highly influence the individual interactions and potentially lead the fuel degradation in non-prototypical directions. The detailed phenomenology of individual interactions and their influence on the transient and on the late phase of the severe accidents are also discussed.

Algoritmos genéticos aplicados a la optimización de antenas Yagi-Uda Genetic algorithms applied to Yagi-Uda antenna optimization

Directory of Open Access Journals (Sweden)

Edgardo César De La Asunción López

2009-07-01

Full Text Available En el presente artículo se muestra un proceso de optimización implementado usando algoritmos genéticos. La población inicial del AG está compuesta por 128 cromosomas con 11 genes por cromosoma. Los cromosomas del AG están compuestos por las longitudes y separaciones de los elementos de la antena Yagi-Uda; los rangos de estos genes fueron escogidos siguiendo estándares de diseño para dichas antenas. Los genes pasan un proceso de análisis para medir cada una las antenas de cada generación de del AG para asignar la aptitud de los individuos. Con el fin de verificar los resultados obtenidos, se aplicaron varias pruebas, entre ellas la construcción de una antena Yagi-Uda optimizada a la cual se le midieron y verificaron sus características electromagnéticas.This paper describes an optimization process implemented using Genetic Algorithms. The initial population of the GA is composed of 128 chromosomes with 11 genes per chromosome. The chromosomes of the GA are composed by the length and separations of the elements of the Yagi-Uda antenna; the ranks of this genes where chosen by design standards for such antennas. All genes undergo a process of analysis to assess every one of the antennas of each generation of the GA to assign the fitness of the individuals. In order to verify the obtained results, various tests were made, and among them excel the construction of the optimized Yagi-Uda antenna to measure and verify it electromagnetic characteristics.

Aplicação do algoritmo de contraponto dissonante de Tenney na determinação de parciais em espectros de sons concretos

Directory of Open Access Journals (Sweden)

Raphael Santos

2014-05-01

Full Text Available Neste artigo, propomos um sistema para determinação de harmonias espectrais extraídas de sons concretos a partir da utilização do algoritmo de contraponto dissonante, desenvolvido por James Tenney. Com base nesse algoritmo, definimos um sistema composicional, descrevemos sua implementação como aplicativo computacional e o utilizamos no planejamento composicional de duas obras originais: o primeiro movimento de um quarteto de clarinetes e um quinteto de metais.

Analysis of BWR out-of-phase instabilities in the frequency domain

International Nuclear Information System (INIS)

Farawila, Y.M.; Pruitt, D.W.; Kreuter, D.

1992-01-01

During startup or because of an inadvertent recirculation pump trip, a boiling water reactor (BWR) may operate at relatively low flow and high power conditions. At these conditions, a BWR is susceptible to coupled flow and power oscillations that could result in undesirable reactor scram unless appropriate countermeasures are taken. This contribution to analytical methods has been developed to address in part a general industrywide and regulatory concern about BWR stability initiated by the LaSalle 2 instability event in March 1988. This work is designed to extend the capability of the one-dimensional parallel channel frequency domain code STAIF to predict the regional oscillation decay ratio. The basic theory follows that developed by March-Leuba and Blakeman, where the oscillation mechanism is identified as the excitation of a subcritical neutronic mode with a constant core pressure drop boundary condition. The improvements to the basic theory include applying the theory to one-dimensional neutronics instead of point kinetics and taking account of the actual three-dimensional harmonic flux distribution

The use of the partial coherence function technique for the investigation of BWR noise dynamics

International Nuclear Information System (INIS)

Kostic, Lj.

1983-01-01

The extensive experimental investigations, at the last time, indicate that the partial coherence function technique can be a powerful method of the investigation of BWR noise dynamics. Symple BWR noise dynamics model for the global noise study, based on different noise phenomena, is proposed in this paper. (author)

Delivering high performance BWR fuel reliably

Energy Technology Data Exchange (ETDEWEB)

Schardt, J.F. [GE Nuclear Energy, Wilmington, NC (United States)

1998-07-01

Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

Estudio de Algoritmos 2-Deslizantes Aplicados al Control de Pilas de Combustible

Directory of Open Access Journals (Sweden)

Cristian Kunusch

2008-07-01

Full Text Available Resumen: En este trabajo se hace un estudio comparativo de tres diferentes técnicas de control por modo deslizante de segundo orden, aplicadas al problema específico del control de respiración de una pila de combustible PEM. Los algoritmos diseñados se contrastan por simulación utilizando el modelo completo del sistema, poniendo particular énfasis en la respuesta transitoria y la robustez frente a perturbaciones. Palabras clave: Pilas de Combustible, Control no lineal, Modo Deslizante

Novel modular natural circulation BWR design and safety evaluation

International Nuclear Information System (INIS)

Ishii, Mamoru; Shi, Shanbin; Yang, Won Sik; Wu, Zeyun; Rassame, Somboon; Liu, Yang

2015-01-01

Highlights: • Introduction of BWR-type natural circulation small modular reactor preliminary design (NMR-50). • Design of long fuel cycle length for the NMR-50. • Design of double passive safety systems for the NMR-50. • RELAP5 analyses of design basis accidents for the NMR-50. - Abstract: The Purdue NMR (Novel Modular Reactor) represents a BWR-type small modular reactor with a significantly reduced reactor pressure vessel (RPV) height. Specifically, it has one third the height of a conventional BWR RPV with an electrical output of 50 MWe. The preliminary design of the NMR-50 including reactor, fuel cycle, and safety systems is described and discussed. The improved neutronics design of the NMR-50 extends the fuel cycle length up to 10 years. The NMR-50 is designed with double passive engineering safety system, which is intended to withstand a prolonged station black out with loss of ultimate heat sink accident such as experienced at Fukushima. In order to evaluate the safety features of the NMR-50, two representative design basis accidents, i.e. main steam line break (MSLB) and bottom drain line break (BDLB), are simulated by using the best-estimate thermal–hydraulic code RELAP5. The RPV water inventory, containment pressure, and the performance of engineering safety systems are investigated for about 33 h after the initiation of the accidents

Application of TRAC-BD1/MOD1 to a BWR/4 feedwater control failure ATWS

International Nuclear Information System (INIS)

Rouhani, S.Z.; Giles, M.M.; Mohr, C.M. Jr.; Weaver, W.L. III.

1984-01-01

This paper begins with a short description of the Transient Reactor Analysis Code for Boiling Water Reactors (TRAC-BWR), briefly mentioning some of its main features such as specific BWR models and input structure. Next, an input model of a BWR/4 is described, and, the assumptions used in performing an analysis of the loss of a feedwater controller without scram are listed. The important features of the calculated trends in flows, pressure, reactivity, and power are shown graphically and commented in the text. A comparison of some of the main predicted trends with the calculated results from a similar study by General Electric is also presented

Automatic segmentation of lizard spots using an active contour model

Directory of Open Access Journals (Sweden)

Jhony Heriberto Giraldo-Zuluaga

2016-01-01

Full Text Available La biometría en animales es una tarea desafiante. En la literatura muchos algoritmos se han utilizado, como, por ejemplo, el reconocimiento de los pechos en pingüinos, el reconocimiento de las orejas en elefantes y el reconocimiento de los patrones de rayas en leopardos. No cabe duda que aún hay mucho trabajo para hacer un uso masivo de la tecnología. En este artículo proponemos un algoritmo de segmentación para extraer manchas de la especie de lagartos Diploglossus millepunctatus, amenazada por la actividad humana. La segmentación automática ha sido lograda con una combinación de preprocesamiento, contornos activos y morfología. Los parámetros de cada etapa del algoritmo de segmentación han sido optimizados usando imágenes de referencia como objetivo. Los resultados muestran que la segmentación automática de manchas es posible. Un 78,37% de segmentación correcta en promedio es alcanzado.

Construction techniques and management methods for BWR plants

International Nuclear Information System (INIS)

Shimizu, Yohji; Tateishi, Mizuo; Hayashi, Yoshishige

1989-01-01

Toshiba is constantly striving for safer and more efficient plant construction to realize high-quality BWR plants within a short construction period. To achieve these aims, Toshiba has developed and improved a large number of construction techniques and construction management methods. In the area of installation, various techniques have been applied such as the modularization of piping and equipment, shop installation of reactor internals, etc. Further, installation management has been upgraded by the use of pre-installation review programs, the development of installation control systems, etc. For commissioning, improvements in commissioning management have been achieved through the use of computer systems, and testing methods have also been upgraded by the development of computer systems for the recording and analysis of test data and the automatic adjustment of controllers in the main control system of the BWR. This paper outlines these construction techniques and management methods. (author)

Paired replacement fuel assemblies for BWR-type reactor

International Nuclear Information System (INIS)

Oguchi, Kazushige.

1997-01-01

There are disposed a large-diameter water rod constituting a non-boiling region at a central portion and paired replacement fuel assemblies for two streams having the same average enrichment degree and different amount of burnable poisons. The paired replacement fuel assemblies comprise a first fuel assembly having a less amount of burnable poisons and a second fuel assembly having a larger amount of burnable poisons. A number of burnable poison-containing fuel rods in adjacent with the large diameter water rod is increased in the second fuel assembly than the first fuel assembly. Then, the poison of the paired replacement fuel assemblies for the BWR type reactor can be annihilated simultaneously at the final stage of the cycle. Accordingly, fuels for a BWR type reactor excellent in economical property and safety and facilitating the design of the replacement reactor core can be obtained. (N.H.)

TLTA/6431, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: The Two-Loop-Test-Apparatus (TLTA) is a 1:624 volume scaled BWR/6 simulator. It was the predecessor of the better-scaled FIST facility. The facility is capable of full BWR system pressure and has a simulated core with a full size 8 x 8, full power single bundle of indirect electrically heated rods. All major BWR systems are simulated including lower plenum, guide tube, core region (bundle and bypass), upper plenum, steam separator, steam dome, annular downcomer, recirculation loops and ECC injection systems. The fundamental scaling consideration was to achieve real-time response. A number of the scaling compromises present in TLTA were corrected in the FIST configuration. These compromises include a number of regional volumes and component elevations. 2 - Description of test: 64.45 sqcm small break LOCA with activation of the full emergency core cooling system, but without activation of the automatic decompression system

Thermohydraulic stability coupled to the neutronic in a BWR

International Nuclear Information System (INIS)

Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A.; Castlllo D, R.

2006-01-01

In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the Laguna Verde

Solving the traffic assignment problem using real data for a segment of Medellin's transportation network

Directory of Open Access Journals (Sweden)

Carlos Alberto González Calderón

2011-01-01

Full Text Available Este artículo está basado en datos obtenidos en los más recientes estudios de transporte que se han realizado en el Área Metropolitana del Valle de Aburrá. (Medellín y otros 9 municipios. Estos estudios fueron la Encuesta Origen Destino (2005, análisis de las rutas de buses (2005 y el Plan Maestro de Movilidad (2006. En el artículo se explica el proceso utilizado para el desarrollo de una aplicación informática para resolver el problema determinístico de equilibrio de usuario en la red vial de Medellín. El código fue construido usando Visual Basic.NET ® y Microsoft Excel ® para la ejecución de algunas operaciones en un segmento de la red vial de Medellín. La distribución del flujo del equilibrio de usuario fue encontrada usando el algoritmo de Frank-Wolfe y fueron analizados algunos aspectos tales como número de iteraciones, patrones de convergencia, tiempo de respuesta y cambios en la demanda de viajes en la red. Los modelos de asignación del tránsito fueron analizados para las horas pico de la tarde. Se compararon los resultados de la asignación del tránsito del algoritmo desarrollado en este trabajo con los resultados de TransCAD ® para los datos del 2005 y fue encontrado que el software es un poco más rápido que el algoritmo, pero sin embargo éste último puede ser una buena herramienta para profesionales y estudiantes para la modelación de redes pequeñas.

BWR and PWR chemistry operating experience and perspectives

International Nuclear Information System (INIS)

Fruzzetti, K.; Garcia, S.; Lynch, N.; Reid, R.

2014-01-01

It is well recognized that proper control of water chemistry plays a critical role in ensuring the safe and reliable operation of Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). State-of-the-art water chemistry programs reduce general and localized corrosion of reactor coolant system, steam cycle equipment, and fuel cladding materials; ensure continued integrity of cycle components; and reduce radiation fields. Once a particular nuclear plant component has been installed or plant system constructed, proper water chemistry provides a global tool to mitigate materials degradation problems, thereby reducing the need for costly repairs or replacements. Recognizing the importance of proper chemistry control and the value in understanding the relationship between chemistry guidance and actual operating experience, EPRI continues to collect, monitor, and evaluate operating data from BWRs and PWRs around the world. More than 900 cycles of valuable BWR and PWR operating chemistry data has been collected, including online, startup and shutdown chemistry data over more than 10 years (> 20 years for BWRs). This paper will provide an overview of current trends in BWR and PWR chemistry, focusing on plants in the U.S.. Important chemistry parameters will be highlighted and discussed in the context of the EPRI Water Chemistry Guidelines requirements (i.e., those parameters considered to be of key importance as related to the major goals identified in the EPRI Guidelines: materials integrity; fuel integrity; and minimizing plant radiation fields). Perspectives will be provided in light of recent industry initiatives and changes in the EPRI BWR and PWR Water Chemistry Guidelines. (author)

Generic safety evaluation report regarding integrity of BWR scram system piping

International Nuclear Information System (INIS)

1981-08-01

Safety concerns associated with postulated pipe breaks in the boiling water reactor (BWR) scram system were identified during the staff's continuing investigation of the Browns Ferry Unit 3 control rod partial insertion failure on June 28, 1980. This report includes an evaluation of the licensing basis for the BWR scram discharge volume (SDV) piping and an assessment of the potential for the SDV piping to fail while in service. A discussion of the means available for mitigation an unlikely SDV system failure is provided. Generic recommendations are made to improve mitigation capability and ensure that system integrity is maintained in service

Repair and preventive maintenance technology for BWR reactor internals and piping

International Nuclear Information System (INIS)

Ootsubo, Tooru; Itou, Takashi; Sakashita, Akihiro

2009-01-01

Stress corrosion cracking of welding portion has found in many domestic and foreign BWR reactor internals and Primary Loop Recirculation piping. Also, repair and preventive maintenance technologies for SCC has been developed and/or adopted to BWRs in recent years. This paper introduces the sample of these technologies, such as seal-welding for SCC on BWR reactor internals, preventive maintenance technology for PLR piping such as Corrosion Resistant Cladding, Internal Polishing and Induction Heating Stress Improvement. These technologies are introduced on 'E-Journal of Advanced Maintenance', which is an international journal on a exclusive website of Japan Society of Maintenology. (author)

BWR type reactors

International Nuclear Information System (INIS)

Yano, Ryoichi; Sato, Takashi; Osaki, Masahiko; Hirayama, Fumio; Watabe, Atsushi.

1980-01-01

Purpose: To effectively eliminate radioactive substances released upon loss of coolant accidents in BWR type reactors. Constitution: A high pressure gas jetting device having a plurality of small aperture nozzles is provided above a spray nozzle, that is, at the top of a dry well. The jetting device is connected to a vacuum breaker provided in a pressure suppression chamber. Upon loss of coolant accident, coolants are sprayed from the spray nozzle and air or nitrogen is jetted from the gas jetting device as well. Then, the gases in the dry well are disturbed, whereby radioactive iodine at high concentration liable to be accumulated in the dry well is forced downwardly, dissolved in the spray water and eliminated. (Ikeda, J.)

Initiation model for intergranular stress corrosion cracking in BWR pipes

International Nuclear Information System (INIS)

Hishida, Mamoru; Kawakubo, Takashi; Nakagawa, Yuji; Arii, Mitsuru.

1981-01-01

Discussions were made on the keys of intergranular stress corrosion cracking of austenitic stainless steel in high-temperature water in laboratories and stress corrosion cracking incidents in operating plants. Based on these discussions, a model was set up of intergranular stress corrosion cracking initiation in BWR pipes. Regarding the model, it was presumed that the intergranular stress corrosion cracking initiates during start up periods whenever heat-affected zones in welded pipes are highly sensitized and suffer dynamic strain in transient water containing dissolved oxygen. A series of BWR start up simulation tests were made by using a flowing autoclave system with slow strain rate test equipment. Validity of the model was confirmed through the test results. (author)

UN ALGORITMO SAEM PARA EL PROBLEMA DE COMPLETACIÓN DE MATRICES // A SAEM ALGORITHM FOR MATRIX COMPLETION PROBLEMS

Directory of Open Access Journals (Sweden)

Anaís Frangeline Acuña Sosa

2015-06-01

Full Text Available In this work we dealt with matrix completion problem. This problem arises in different fields, for example, systems and control theory, image processing and collaborative filtering. Given a probabilistic matrix factorization model, we present an approach based on Bayesian statistics and a stochastic expectation maximization algorithm to retrieve an array of data from a sample of its inputs. The proposed method does not require regularization parameters and estimates the rank of the matrix, in contrast to the BPMF method. Our results show that the proposed method outperforms to an augmented lagrangian algorithm and the BPMF method in its ability to find the rank of the matrix and in efficiency respectively. // RESUMEN En este trabajo estudiamos el problema de completación de matrices. Este problema se presenta en diversas áreas como la teoría de sistemas y control, procesamiento de imágenes y filtrado colaborativo. Considerando un modelo de factorización probabilística de matrices, establecemos una propuesta basada en estadística Bayesiana y un algoritmo EM estocástico para recubrir una matriz de datos a partir de una muestras de sus entradas. El método propuesto no requiere de parámetros de regularización y da un estimado del rango de la matriz, en contraste con el método BPMF. Los resultados muestran que el algoritmo propuesto da mejores estimados del rango de la matriz en comparación con un algoritmo basado en lagrangeanos aumentados y es más eficiente que el método BPMF.

Validation and application of the system code ATHLET-CD for BWR severe accident analyses

Energy Technology Data Exchange (ETDEWEB)

Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Imke, Uwe; Sanchez, Victor

2016-10-15

Highlights: • We present the application of the system code ATHLET-CD code for BWR safety analyses. • Validation of core in-vessel models is performed based on KIT CORA experiments. • A SB-LOCA scenario is simulated on a generic German BWR plant up to vessel failure. • Different core reflooding possibilities are investigated to mitigate the accident consequences. • ATHLET-CD modelling features reflect the current state of the art of severe accident codes. - Abstract: This paper is aimed at the validation and application of the system code ATHLET-CD for the simulation of severe accident phenomena in Boiling Water Reactors (BWR). The corresponding models for core degradation behaviour e.g., oxidation, melting and relocation of core structural components are validated against experimental data available from the CORA-16 and -17 bundle tests. Model weaknesses are discussed along with needs for further code improvements. With the validated ATHLET-CD code, calculations are performed to assess the code capabilities for the prediction of in-vessel late phase core behaviour and reflooding of damaged fuel rods. For this purpose, a small break LOCA scenario for a generic German BWR with postulated multiple failures of the safety systems was selected. In the analysis, accident management measures represented by cold water injection into the damaged reactor core are addressed to investigate the efficacy in avoiding or delaying the failure of the reactor pressure vessel. Results show that ATHLET-CD is applicable to the description of BWR plant behaviour with reliable physical models and numerical methods adopted for the description of key in-vessel phenomena.

Boiling transition phenomenon in BWR fuel assemblies effect of fuel spacer shape on critical power

International Nuclear Information System (INIS)

Yamamoto, Yasushi; Morooka, Shin-ichi; Mitsutake, Toru; Yokobori, Seiichi; Kimura, Jiro.

1996-01-01

A thorough understanding of the thermal-hydraulic phenomena near fuel spacer is necessary for the accurate prediction of the critical power of BWR fuel assemblies, and is thus essential for effective developments of a new BWR fuel assembly. The main purpose of this study is to develop an accurate method for predicting the effect of spacer shapes on critical power. Tests have been conducted under actual BWR operating conditions, using an annulus flow channel consisting of a heated rod and circular-tube channel, and BWR simulated 4x4 rod bundles with heater rods unheated just upsteam of spacer. The effect of spacer shapes on critical power was predicted analytically based on the droplet deposition rate estimation. The droplet deposition rate for different spacer shapes was calculated using a single-phase flow model. The prediction results were compared with the test results for the annulus flow channel using ring-type spacers. Analytical results of critical power agreed with measured critical power from point of the effects of changes in the rod-spacer clearance and the spacer thickness on critical power. (author)

Efecto del tamaño del alfabeto en el rendimiento de un algoritmo de compresión probabilístico

Directory of Open Access Journals (Sweden)

Carlos Rincón

2012-12-01

Full Text Available El presente trabajo tuvo como finalidad determinar el efecto del tamaño del alfabeto de un mensaje, en el rendimiento del algoritmo de compresión probabilístico basado en la posición de los símbolos, el cual propusimos en un trabajo previo (Rincón, Acurero, Bracho y Jakymec, 2008. La metodología utilizada consistió en 7 etapas: (a determinación de las variables dependientes e independientes a objeto de estudio, (b desarrollo e implementación del algoritmo propuesto, (c construcción de los archivos de prueba, (d ejecución del algoritmo implementado sobre los archivos de prueba, (e determinación del modelo matemático que explique el comportamiento de las variables dependientes, (f aplicación del método estadístico análisis de varianza, (g análisis de los resultados obtenidos. Las variables dependientes seleccionadas fueron el tiempo de compresión y la relación de compresión. El diseño del modelo estadístico seleccionado fue un totalmente aleatorizado con tratamiento en un arreglo factorial 4x2, con dos factores: tamaño del alfabeto (4,8,12 y 16 símbolos y distribucción probabilística del alfabeto (aleatorio y equiprobable. Del análisis de varianza se obtuvo diferencias significativas para todas las variables independientes y su interacción en todas las variables dependientes, corroborando así el efecto que tiene el tamaño del alfabeto en el rendimiento del algoritmo de compresión estudiado. La prueba de Tukey determinó que para la variable tiempo de compresión el mejor rendimiento se obtiene con la distribución aleatoria y el mayor tamaño del alfabeto (12 y 16, mientras que para la variable relación de compresión, el mejor rendimiento se obtiene con la distribución aleatoria y el menor tamaño del alfabeto.

Prediction of BWR performance under the influence of Isolation Condenser-using RAMONA-4 code

International Nuclear Information System (INIS)

Khan, H.J.; Cheng, H.S.; Rohatgi, U.S.

1992-01-01

The purpose of the Boiling Water Reactor (BWR) Isolation Condenser (IC) is to passively control the reactor pressure by removing heat from the system. This type of control is expected to reduce the frequency of opening and closing of the Safety Relief Valves (SRV). A comparative analysis is done for a BWR operating with and without the influence of an IC under Main Steam Isolation Valve (MSIV) closure. A regular BWR, with forced flow and high thermal power, has been considered for analysis. In addition, the effect of ICs on the BWR performance is studied for natural convection flow at lower power and modified riser geometry. The IC is coupled to the steam dome for the steam inlet flow and the Reactor Pressure Vessel (RPV) near the feed water entrance for the condensate return flow. Transient calculations are performed using prescribed pressure set points for the SRVs and given time settings for MSIV closure. The effect of the IC on the forced flow is to reduce the rate of pressure rise and thereby decrease the cycling frequency ofthe SRVS. This is the primary objective of any operating IC in a BWR (e.g. Oyster Creek). The response of the reactor thermal and fission power, steam flow rate, collapsed liquid level, and core average void fraction are found to agree with the trend of pressure. The variations in the case of an active IC can be closely related to the creation of a time lag and changes in the cycling frequency of the SRVS. An analysis for natural convection flow in a BWR indicates that the effect of an IC on its transient performance is similar to that for the forced convection system. In this case, the MSIV closure, has resulted in a lower peak pressure due to the magnitude of reduced power. However, the effect of reduced cycling frequency of the SRV due to the IC, and the time lag between the events, are comparable to that for forced convection

Subchannel analysis of a critical power test, using simulated BWR 8x8 fuel assembly

International Nuclear Information System (INIS)

Mitsutake, T.; Terasaka, H.; Yoshimura, K.; Oishi, M.; Inoue, A.; Akiyama, M.

1990-01-01

Critical power predictions have been compared with the critical power test data obtained in simulated BWR 8x8 fuel rod assemblies. Two analytical methods for the critical power prediction in rod assemblies are used in the prediction, which are the subchannel analysis using the COBRA/BWR subchannel computer code with empirical critical heat flux (CHF) correlations and the liquid film dryout estimation using the CRIPP-3F 'multi-fluid' computer code. Improvements in both the analytical methods were made for spacer effect modeling, though they were specific for application to the current BWR rod assembly type. In general a reasonable agreement was obtained, though comparisons, between the prediction and the obtained test data. (orig.)

The impact of BWR MK I primary containment failure dynamics on secondary containment integrity

International Nuclear Information System (INIS)

Greene, S.R.

1987-01-01

During the past four years, the ORNL BWRSAT Program has developed a series of increasingly sophisticated BWR secondary containment models. These models have been applied in a variety of studies to evaluate the severe accident mitigation capability of BWR secondary containments. This paper describes the results of a recent ORNL study of the impact of BWR MK I primary containment failure dynamics on secondary containment integrity. A 26-cell MELCOR Browns Ferry secondary containment model is described and the predicted thermodynamic response of the secondary containment to a variety of postulated primary containment failure modes is presented. The effects of primary containment failure location, timing, and ultimate hole size on secondary containment response is investigated, and the potential impact of hydrogen deflagrations on secondary containment integrity is explored

BWR internals life assurance

International Nuclear Information System (INIS)

Herrera, M.L.; Stancavage, P.P.

1988-01-01

Boiling water reactor (BWR) internal components play an important role in power plant life extension. Many important internals were not designed for easy removal and changes in material properties and local environmental effects due to high radiation makes stress corrosion cracking more likely and more difficult to correct. Over the past several years, operating experience has shown that inspection, monitoring and refurbishment can be accomplished for internal structures with existing technology. In addition, mitigation techniques which address the causes of degradation are available to assure that life extension targets can be met. This paper describes the many considerations and aspects when evaluating life extension for reactor vessel internals

Damage by radiation in structural materials of BWR reactor vessels; Dano por radiacion en materiales estructurales de vasijas de reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

2002-07-01

The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA Mark III Salazar reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A{sup 2}). (Author)

Eulerian fluid-structure analysis of BWR

International Nuclear Information System (INIS)

McMaster, W.H.

1979-05-01

A fluid-structure-interaction algorithm is developed for the analysis of the dynamic response of a BWR pressure-suppression pool and containment structure. The method is incorporated into a two-dimensional semi-implicit Eulerian hydrodynamics code, PELE-IC, for the solution of incompressible flow coupled to flexible structures. The fluid, structure, and coupling algorithms have been verified by calculation of solved problems from the literature and by comparison with air and steam blowdown experiments

ZZ BWRSB-RINGHALS1, Stability Benchmark Data from BWR RINGHALS-1

International Nuclear Information System (INIS)

2002-01-01

Description of program or function: The purpose of this benchmark is to enable code developers to test their codes and also to validate the predictive capability of their respective codes and models for BWR stability analysis. Emphasis is put on the modelling of flow dynamics of the reactor core and in-vessel flow loop wit detailed neutronic and thermodynamic feedback. The secondary systems as well as the control and production systems will be neglected. Data provided comes from measurements in beginning of cycle (BOC) 14, 15, 16 and 17 and middle of cycle (MOC) 16 in the Swedish BWR reactor Ringhals 1. For these measurements complete data sets are given

BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

International Nuclear Information System (INIS)

Chiang, Ren-Tai; Williams, John B.; Folk, Ken S.

2008-01-01

MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)

BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

Energy Technology Data Exchange (ETDEWEB)

Chiang, Ren-Tai [University of Florida, Gainesville, Florida 32611 (United States); Williams, John B.; Folk, Ken S. [Southern Nuclear Company, Birmingham, Alabama 35242 (United States)

2008-07-01

MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)

Product Evaluation Task Force Phase Two report for BWR/PWR dissolver wastes

International Nuclear Information System (INIS)

Francis, A.J.

1990-01-01

It has been proposed that all Intermediate Level Wastes arising at Sellafield should be encapsulated prior to ultimate disposal. The Product Evaluation Task Force (PETF) was set up to investigate possible encapsulants and to produce an adequate data base to justify the preferred matrices. This report details the work carried out, under Phase 2 of the Product Evaluation Task Force programme, on BWR/PWR Dissolver Wastes. Three possible types of encapsulants for BWR/PWR Dissolver Wastes:- Inorganic cements, Polymer cements and Polymers are evaluated using the Kepner Tregoe decision analysis technique. This technique provides a methodology for scoring and ranking alternative options and evaluating any risks associated with an option. The analysis shows that for all four stages of waste management operations ie Storage, Transport, handling and emplacement, Disposal and Process, cement matrices are considerably superior to other potential matrices. A matrix, consisting of three parts Blast Furnace Slag (BFS) to one part Ordinary Portland Cement (OPC), is recommended for Phase 3 studies on BWR/PWR Dissolver Wastes. (author)

System control model of a turbine for a BWR; Modelo del sistema de control de una turbina para un BWR

Energy Technology Data Exchange (ETDEWEB)

Vargas O, Y. [Universidad del Valle de Mexico, Campus Toluca, Av. Las Palmas No. 136, Col. San Jorge Pueblo Nuevo, 52140 Metepec, Estado de Mexico (Mexico); Amador G, R.; Ortiz V, J.; Castillo D, R.; Delfin L, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: rodolfo.amador@inin.gob.mx

2009-10-15

In this work is presented a design of a control system of a turbine for a nuclear power plant with a BWR like energy source. The model seeks to implement later on at thermal hydraulics code of better estimate RELAP/SCDAPSIM. The model is developed for control and protection of turbine, and the consequent protection to the BWR, considering that the turbine control could be employed for one or several turbines in series. The quality of present designs of control pattern of turbine it is that it considers the parameters more important in the operation of a turbine besides that is has incorporated at control the secondary parameters that will be activated originally as true when the turbine model is substituted by a model more detailed. The development of control model of a turbine will be good in short and medium term to realize analysis about the operation of turbine with different operation conditions, of vapor extraction specific steps of turbine to feed other equipment s, besides analyzing the separate effect and integrated effect. (Author)

Studies of fragileness in steels of vessels of BWR reactors; Estudios de fragilizacion en aceros de vasija de reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2003-07-01

The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA MARK lll reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A{sup 2}. (Author)

Algorithms for the evaluation of the reliability of networks and substations; Algoritmos para evaluar la confiabilidad de redes y subestaciones electricas

Energy Technology Data Exchange (ETDEWEB)

Vega Ortiz, Miguel [Instituto de Investigaciones Electricas, Cuernavaca, Morelos (Mexico)

2001-07-01

Two algorithms are presented, one for analyzing the reliability of networks and another one to analyzing the reliability of electrical substations. The algorithms model a network or a substation by means of a graph and determine all the exits, of up to third degree, which isolates the load points. The algorithm that is used for the network analysis, is the one of p-cuts and in the case of the substations it is the one of minimum ways, for later, from these minimum p-cuts and of the minimum ways, obtains the minimum cuts. These minimum cuts are evaluated with the approximate method and give as a result the system reliability and of each one of the components that have been modeled. At the end some examples and results obtained are shown. As case of study, the algorithm is applied to two substations arrangements and to a network of industrial distribution, but it is possible to be applied to transmission and distribution networks. [Spanish] Se presentan dos algoritmos, uno para analizar la confiabilidad de redes y otro para analizar la confiabilidad de subestaciones electricas. Los algoritmos modelan una red o una subestacion por medio de un grafo y determinan todas las salidas, de hasta tercer grado, que aislan los puntos de carga. El algoritmo que se utiliza para el analisis de redes, es el de p-cortes y en el caso de las subestaciones es el de caminos minimos, para despues, a partir de estos p-cortes minimos y de los caminos minimos, obtener los cortes minimos. Estos cortes minimos se evaluan con el metodo aproximado y dan como resultado la confiabilidad del sistema y de cada una de las componentes que se hayan modelado. Al final se muestran algunos ejemplos y resultados obtenidos. Como caso de estudio, el algoritmo se aplica a dos arreglos de subestaciones y a una red de distribucion industrial, pero se puede aplicar a redes de transmision y de distribucion.

GENERALIZACIÓN ALGEBRAICA DE LA DFC: REFLEXIONES POR MEDIO DE UN ALGORITMO ALGEBRAICO

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Antonio Geloneze Neto

2012-09-01

Full Text Available La Demostración de Flujos de Caja (DFC pasó a ser un informe obligatorio por la contabilidad a partir del 1 de enero de 2008 para todas las empresas de capital abierto o con patrimonio líquido superior a dos millones de rea¬les y, de esa forma, se torna otro importante informe para la toma de decisiones gerenciales. Este trabajo tiene por objetivo proponer una generalización algébrica para la DFC. Papeles de trabajo pueden contribuir para cerrar una laguna didáctica en la enseñanza de la DFC y producir el método indirecto y el método directo, lado a lado con su equivalencia destacada, en una misma matriz por medio de algoritmos algebraicos. La pesquisa es de naturale¬za normativa y enfatiza el carácter transversal entre la Contabilidad y la Matemática, mostrando que los informes contables y sus estructuras pueden ser vistos como matrices y sujetos a deducciones algebraicas sobre los eventos registrados por medio de las partidas dobles. Como resultado, se pudo demostrar un algoritmo matemático con matrices y sub-matrices y un guión en el formato de papeles de trabajo, compatibles con las orientaciones norma¬tivas para la DFC en la legislación brasileña, que permite una DFC clara, segura y efectiva.

Prevention of organic iodide formation in BWR's

International Nuclear Information System (INIS)

Karjunen, T.; Laitinen, T.; Piippo, J.; Sirkiae, P.

1996-01-01

During an accident, many different forms of iodine may emerge. Organic iodides, such as methyl iodide and ethyl iodide, are relatively volatile, and thus their appearance leads to increased concentration of gaseous iodine. Since organic iodides are also relatively immune to most accident mitigation measures, such as sprays and filters, they can affect the accident source term significantly even when only a small portion of iodine is in organic form. Formation of organic iodides may not be limited by the amount of organic substances available. Excessive amounts of methane can be produced, for example, during oxidation of boron carbide, which is used in BWR's as a neutron absorber material. Another important source is cable insulation. In a BWR, a large quantity of cables is placed below the pressure vessel. Thus a large quantity of pyrolyse gases will be produced, should the vessel fail. Organic iodides can be formed as a result of many different reactions, but at least in certain conditions the main reaction takes place between an organic radical produced by radiolysis and elemental iodine. A necessary requirement for prevention of organic iodide production is therefore that the pH in the containment water pools is kept high enough to eliminate formation of elemental iodine. In a typical BWR the suppression pool water is usually unbuffered. As a result, the pH may be dominated by chemicals introduced during an accident. If no system for adding basic chemicals is operable, the main factor affecting pool water pH may be hydrochloric acid released during cable degradation. Should this occur, the conditions could be very favorable for production of elemental iodine and, consequently, formation of organic iodides. Although high pH is necessary for iodine retention, it could have also adverse effects. High pH may, for example, accelerate corrosion of containment materials and alter the characteristics of the solid corrosion products. (author) 6 figs., 1 tab., 13 refs

Prony's method application for BWR instabilities characterization

Energy Technology Data Exchange (ETDEWEB)

Castillo, Rogelio, E-mail: rogelio.castillo@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Ramírez, J. Ramón, E-mail: ramon.ramirez@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Alonso, Gustavo, E-mail: gustavo.alonso@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Instituto Politecnico Nacional, Unidad Profesional Adolfo Lopez Mateos, Ed. 9, Lindavista, D.F. 07300 (Mexico); Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico)

2015-04-01

Highlights: • Prony's method application for BWR instability events. • Several BWR instability benchmark are assessed using this method. • DR and frequency are obtained and a new parameter is proposed to eliminate false signals. • Adequate characterization of in-phase and out-of-phase events is obtained. • The Prony's method application is validated. - Abstract: Several methods have been developed for the analysis of reactor power signals during BWR power oscillations. Among them is the Prony's method, its application provides the DR and the frequency of oscillations. In this paper another characteristic of the method is proposed to determine the type of oscillations that can occur, in-phase or out-of-phase. Prony's method decomposes a given signal in all the frequencies that it contains, therefore the DR of the fundamental mode and the first harmonic are obtained. To determine the more dominant pole of the system a normalized amplitude W of the system is calculated, which depends on the amplitude and the damping coefficient. With this term, it can be analyzed which type of oscillations is present, if W of the fundamental mode frequency is the greater, the type of oscillations is in-phase, if W of the first harmonic frequency is the greater, the type of oscillations is out-of-phase. The method is applied to several stability benchmarks to assess its validity. Results show the applicability of the method as an alternative analysis method to determine the type of oscillations occurred.

Prony's method application for BWR instabilities characterization

International Nuclear Information System (INIS)

Castillo, Rogelio; Ramírez, J. Ramón; Alonso, Gustavo; Ortiz-Villafuerte, Javier

2015-01-01

Highlights: • Prony's method application for BWR instability events. • Several BWR instability benchmark are assessed using this method. • DR and frequency are obtained and a new parameter is proposed to eliminate false signals. • Adequate characterization of in-phase and out-of-phase events is obtained. • The Prony's method application is validated. - Abstract: Several methods have been developed for the analysis of reactor power signals during BWR power oscillations. Among them is the Prony's method, its application provides the DR and the frequency of oscillations. In this paper another characteristic of the method is proposed to determine the type of oscillations that can occur, in-phase or out-of-phase. Prony's method decomposes a given signal in all the frequencies that it contains, therefore the DR of the fundamental mode and the first harmonic are obtained. To determine the more dominant pole of the system a normalized amplitude W of the system is calculated, which depends on the amplitude and the damping coefficient. With this term, it can be analyzed which type of oscillations is present, if W of the fundamental mode frequency is the greater, the type of oscillations is in-phase, if W of the first harmonic frequency is the greater, the type of oscillations is out-of-phase. The method is applied to several stability benchmarks to assess its validity. Results show the applicability of the method as an alternative analysis method to determine the type of oscillations occurred

PROCESAMIENTO DE IMÁGENES PARA LA PLANEACIÓN DE RUTAS PARA ROBOTS MÓVILES BASADOS EN LEGO MINDSTORMS NXT

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Jesús Alberto Torres Velásquez

2014-11-01

Full Text Available El presente artículo contiene el procedimiento para implementar diferentes técnicas de procesamiento de  imágenes y principalmente el algoritmo de esqueletización con el objetivo de planificar rutas libres de obstáculos para robots móviles; mediante una fotografía tomada sobre un área determinada con los obstáculos y el robot (construido con el kit de robótica LEGO MINDSTORMS NXT. Se realiza el debido procesamiento a dicha imagen, para luego, de forma remota enviar una serie de órdenes vía Bluetooth desde un computador usando MATLAB; ordenes generadas por el algoritmo aplicado sobre la imagen ya procesada, con el fin de obtener el desplazamiento de la plataforma móvil a un punto de llegada determinado por el usuario. 

Generalización algebraica de la DFC: reflexiones por medio de un algoritmo algebraico

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Antonio Geloneze Neto

2012-01-01

Full Text Available La Demostración de Flujos de Caja (DFC pasó a ser un informe obligatorio por la contabilidad a partir del 1 deenero de 2008 para todas las empresas de capital abierto o con patrimonio líquido superior a dos millones de realesy, de esa forma, se torna otro importante informe para la toma de decisiones gerenciales. Este trabajo tiene porobjetivo proponer una generalización algébrica para la DFC. Papeles de trabajo pueden contribuir para cerrar unalaguna didáctica en la enseñanza de la DFC y producir el método indirecto y el método directo, lado a lado consu equivalencia destacada, en una misma matriz por medio de algoritmos algebraicos. La pesquisa es de naturalezanormativa y enfatiza el carácter transversal entre la Contabilidad y la Matemática, mostrando que los informescontables y sus estructuras pueden ser vistos como matrices y sujetos a deducciones algebraicas sobre los eventosregistrados por medio de las partidas dobles. Como resultado, se pudo demostrar un algoritmo matemático conmatrices y sub-matrices y un guión en el formato de papeles de trabajo, compatibles con las orientaciones normativaspara la DFC en la legislación brasileña, que permite una DFC clara, segura y efectiva.

BWR-stability investigation at Forsmark 1

International Nuclear Information System (INIS)

Bergdahl, B.G.; Reisch, F.; Oguma, R.; Lorenzen, J.; Aakerhielm, F.

1988-01-01

A series of noise measurements have been conducted at Forsmark 1 during start-up operation after the revision summer '87. The main purpose was to investigate BWR-stability problems, i.e. resonant power oscillations of 0.5 Hz around 65% power and 4100 kg/s core flow, which tend to arise at high power and low core flow conditions. The analysis was performed to estimate the noise source which gives rise to the oscillation, to evaluate the measure of stability, i.e. the Decay Ratio (Dr) as well as to investigate other safety related problems. The result indicates that the oscillation is due to the dynamic coupling between the neutron kinetics and thermal hydraulics via void reactivity feedback. The Dr ranged between values of 0.7 and > 0.9, instead of expected 0.6 (Dr=1 is defined as instability). These high values imply that the core cannot suppress oscillations fast enough and a small perturbation can cause scram. Further it was found that the entire core is oscillating in phase (LPRM's) with varying strength where any connection to the consequences of different fuel (8x8, 9x9) being present simultaneously cannot be excluded. This report elucidates the importance of an on-line BWR-stability surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

Identification and assessment of BWR in-vessel severe accident mitigation strategies

Energy Technology Data Exchange (ETDEWEB)

Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M. [Oak Ridge National Lab., TN (United States)

1992-10-01

This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners` Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored.

Identification and assessment of BWR in-vessel severe accident mitigation strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M.

1992-10-01

This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

BWR 9 X 9 Fuel Assembly Thermal-Hydraulic Tests (2): Hydraulic Vibration Test

International Nuclear Information System (INIS)

Yoshiaki Tsukuda; Katsuichiro Kamimura; Toshiitsu Hattori; Akira Tanabe; Noboru Saito; Masahiko Warashina; Yuji Nishino

2002-01-01

Nuclear Power Engineering Corporation (NUPEC) conducted thermal-hydraulic projects for verification of thermal-hydraulic design reliability for BWR high-burnup 8 x 8 and 9 x 9 fuel assemblies, entrusted by the Ministry of Economy, Trade and Industry (METI). As a part of the NUPEC thermal-hydraulic projects, hydraulic vibration tests using full-scale test assemblies simulating 9 x 9 fuel assemblies were carried out to evaluate BWR fuel integrity. The test data were applied to development of a new correlation for the estimation of fuel rod vibration amplitude. (authors)

Standard Technical Specifications, General Electric Plants, BWR/6

International Nuclear Information System (INIS)

1992-09-01

This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

BWR power oscillation evaluation methodologies in core design

International Nuclear Information System (INIS)

Hotta, Akitoshi

1995-01-01

At the initial stage of BWR development, the power oscillation due to the nuclear-thermal interaction originated in random boiling phenomena and nuclear void feedback was feared. But it was shown that under the high pressure condition in the normal operation of recent commercial BWRs, the core is in very stable state. However, power oscillation events have been observed in actual machines, and it is necessary to do the stability evaluation that sufficiently reflects the detailed operation conditions of actual plants. As the cause of power oscillation events, the instability of control system and nuclear-thermal coupling instability are important, and their mechanisms are explained. As the model for analyzing the stability of BWR core, the nuclear-thermal coupling model in frequency domain is the central existence. As the information for the design, the parameters of fuel assemblies, and the nuclear parameters and the thermohydraulic parameters of cores are enumerated. LAPUR-TSI is a nuclear-thermal coupling model. The analysis system in the software of Tokyo Electric Power Co. is outlined, and the analysis model was verified. (K.I.)

Standard Technical Specifications, General Electric plants, BWR/4

International Nuclear Information System (INIS)

1992-09-01

This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

Un algoritmo basado en búsqueda tabú granular para la solución de un problema de ruteo de vehículos considerando flota heterogénea

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Duván Eduardo Puenayán

2014-12-01

Full Text Available Este artículo aborda el problema de ruteo de vehículos con flota heterogénea (VRPH, en el cual se busca determinar las rutas a ser construidas para satisfacer las demandas de los clientes, considerando una flota de vehículos con capacidad y costos no homogéneos. El objetivo es minimizar la distancia total de las rutas recorridas por los diferentes vehículos. En este artículo, se propone un algoritmo metaheurístico basado en una búsqueda tabú granular para la solución del problema. El algoritmo acepta soluciones infactibles penalizadas por un factor dinámico que se ajusta durante la búsqueda. Experimentos computacionales en instancias reales de una compañía colombiana muestran que el algoritmo propuesto es capaz de obtener, en tiempos computacionales reducidos, mejores soluciones que las obtenidas por el método tradicional de planificación de rutas, usado en la compañía.

Un nuevo algoritmo para la optimación de estructuras: el recocido simulado

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Vázquez Espí, Mariano

1995-04-01

método de resolución, salvo para ciertos casos particulares en los que la función y sus variables satisfacen condiciones específicas (continuidad, diferenciabilidad, etc.. En numerosos casos de índole práctica no se dan tales condiciones y, en consecuencia, la solución al problema se aproxima, existiendo para ello una variedad de métodos heurísticos. El recocido simulado es un algoritmo de aproximación a la solución óptima, fundado en una analogía con el comportamiento de sistemas termodinámicos simples y viene siendo utilizado en ciertos problemas de ingeniería. El presente trabajo muestra cómo usarlo en la teoría de estructuras, señalando sus ventajas (universalidad así como sus inconvenientes (lentitud. Para ello, se describen brevemente los distintos problemas de la teoría de estructuras, así como aspectos fundamentales de la teoría de algoritmos. Con mayor detalle, se describe la forma matemática del algoritmo de recocido. En ambos casos se emplea un problema paradigmático de optimación: el del viajante de comercio. Finalmente, se muestra el empleo del algoritmo para "aproximar" formas óptimas de cerchas isostáticas.

Desempeño de Algoritmos Gusano en la Navegación de una Plataforma Robótica Móvil

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Luis Alvarez

2013-11-01

Full Text Available El presente trabajo presenta la implementación de algoritmos de navegación denominados gusano 1 y gusano 2 sobre la plataforma robotica móvil de iRobot Create, los cuales se brindaron la capacidad de autonomía para navegar dentro de entornos desconocidos. Además, se utilizaron los módulos de radio frecuencia, que funcionan bajo el estándar ZigBee, para dotar de comunicación inalambrica entre el robot y una PC. Se realizaron pruebas de comunicación para determinar la máxima distancia de trabajo a la que pueden funcionar los módulos RF en interiores; mientras que, para el análisis de desempeño de los algoritmos básicos de navegación montados sobre la plataforma robótica móvil se definieron tres escenarios diferentes donde se verificó la certeza de haber alcanzado la meta por el robot.

Impacto de la memoria cache en la aceleración de la ejecución de algoritmo de detección de rostros en sistemas empotrados

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Alejandro Cabrera Aldaya

2012-06-01

Full Text Available En este trabajo se analiza el impacto de la memoria cache sobre la aceleración de la ejecución del algoritmo de detección de rostros de Viola-Jones en un sistema de procesamiento basado en el procesador Microblaze empotrado en un FPGA. Se expone el algoritmo, se describe una implementación software del mismo y se analizan sus funciones más relevantes y las características de localidad de las instrucciones y los datos. Se analiza el impacto de las memorias cache de instrucciones y de datos, tanto de sus capacidades (entre 2 y 16 kB como de tamaño de línea (de 4 y 8 palabras. Los resultados obtenidos utilizando una placa de desarrollo Spartan3A Starter Kit basada en un FPGA Spartan3A XC3S700A, con el procesador Microblaze a 62,5 MHz y 64 MB de memoria externa DDR2 a 125 MHz,  muestran un mayor impacto de la cache de instrucciones que la de datos, con valores óptimos de 8kB para la cache de instrucciones y entre 4 y 16kB para la cache de datos. Con estas memorias se alcanza una aceleración de 17 veces con relación a la ejecución del algoritmo en memoria externa. El tamaño de la línea de cache tiene poca influencia sobre la aceleración del algoritmo.

MELCOR 1.8.2 assessment: The DF-4 BWR Damaged Fuel experiment

International Nuclear Information System (INIS)

Tautges, T.J.

1993-10-01

MELCOR is a fully integrated, engineering-level computer code being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As a part of an ongoing assessment, program, MELCOR has been used to model the ACRR in-pile DF-4 Damaged Fuel experiment. DF-4 provided data for early phase melt progression in BWR fuel assemblies, particularly for phenomena associated with eutectic interactions in the BWR control blade and zircaloy oxidation in the canister and cladding. MELCOR provided good agreement with experimental data in the key areas of eutectic material behavior and canister and cladding oxidation. Several shortcomings associated with the MELCOR modeling of BWR geometries were found and corrected. Twenty-five sensitivity studies were performed on COR, HS and CVH parameters. These studies showed that the new MELCOR eutectics model played an important role in predicting control blade behavior. These studies revealed slight time step dependence and no machine dependencies. Comparisons made with the results from four best-estimate codes showed that MELCOR did as well as these codes in matching DF-4 experimental data

BWR ATWS mitigation by Fine Motion Control Rod

International Nuclear Information System (INIS)

Rohatgi, U.S.; Cheng, H.S.; Khan, H.; Mallen, A.; Diamond, D.

1994-01-01

Two main methods of ATWS mitigation in a SBWR are: fine Motion control Rods (FMCRD) and Boron injection via the Standby Liquid control System (SLCS). This study has demonstrated that the use of FMCRD along with feedwater runback mitigated the conditions due to reactivity insertion and possible ATWS in a BWR which is similar to SBWR

RECONFIGURACIÓN DE REDES ELÉCTRICAS DE MEDIA TENSIÓN BASADA EN EL ALGORITMO DE PRIM RECONFIGURATION OF MEDIUM VOLTAGE NETWORKS BASED ON PRIM'S ALGORITHM

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Angely Cárcamo-Gallardo

2007-04-01

Full Text Available En este trabajo se presenta un nuevo algoritmo que permite reconfigurar un sistema de distribución (SD de energía eléctrica minimizando la energía no suministrada (ENS. El SD se modela utilizando teoría de grafos, mientras que la ENS se formula recursivamente y se parametriza en términos de los índices de confiabilidad del SD. Empleando esta modelación se transforma el problema de optimización en el problema de encontrar el árbol de mínima expansión (AME a partir del grafo que modela al SD, donde la métrica de distancia utilizada corresponde a la ENS a cada nodo del SD. Para encontrar de manera eficiente el AME se utiliza el algoritmo de Prim, ya que pertenece a la clase de algoritmos voraces en el cálculo del AME. Adicionalmente, se propone un algoritmo que realiza una revisión del AME obtenido analizando las topologías que fueron descartadas aleatoriamente durante el proceso de decisión. El desempeño del algoritmo de optimización se evalúa en sistemas de pruebas y en dos sistemas eléctricos reales.This paper presents a novel algorithm to reconfigure an electric power distribution network (EPDN, minimizing its non-supplied energy (NSE. The EPDN is modeled using graph theory and the NSE is recursively formulated in terms of the reliability parameters of the EPDN. Based on this mathematical model, we transform the original optimization problem into the graph theory problem of finding the minimum spanning tree (MST of a given graph, which models the EPDN. The distance metric employed by the searching algorithm is the NSE. In order to efficiently find the MST, Prim's algorithm is employed due to is greedy search behavior. In addition, a backtracking algorithm is used to check the MST obtained. The backtracking algorithm analyzes all the candidate topologies that were randomly discarded during the decision process. The performance of the optimization algorithm is evaluated using testing systems and two actual EPDNs.

The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

International Nuclear Information System (INIS)

Mursid Djokolelono.

1976-01-01

Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

Reconstrucción filogenética usando geometría algebraica

Directory of Open Access Journals (Sweden)

Casanellas, Marta

2010-12-01

Full Text Available A new approach to phylogenetic reconstruction has been emerging in the last years. Given an evolutionary model, the joint probability distribution of the nucleotides for these species satisfy some algebraic constraints called invariants. These invariants have theoretical and practical interest, since they can be used to infer phylogenies. In this paper, we explain how to use these invariants to design algorithms for phylogenetic reconstruction and we show how the application of tools and theoretical results coming from commutative algebra and algebraic geometry can improve the performance and the efficiency of these algorithms.

Una nueva aproximación a la reconstrucción filogenética basada en la geometría algebraica está ganando fuerza en los últimos años. Fijado un modelo evolutivo para un conjunto de especies, las distribuciones teóricas de los nucleótidos de estas especies satisfacen ciertas relaciones algebraicas que llamamos invariantes. Estos invariantes son de interés teórico y práctico dado que se pueden utilizar para inferir filogenias. En este artículo, explicamos cómo usar los invariantes para implementar algoritmos de reconstrucción filogenética y mostramos cómo el uso de técnicas y resultados teóricos procedentes del álgebra conmutativa y la geometría algebraica puede contribuir en la mejora en la eficacia y la eficiencia de estos algoritmos.

High burnup (41 - 61 GWd/tU) BWR fuel behavior under reactivity initiated accident conditions

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Takehiko; Kusagaya, Kazuyuki; Yoshinaga, Makio; Uetsuka, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-12-01

High burnup boiling water reactor (BWR) fuel was pulse irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity initiated accident (RIA) conditions. Temperature, deformation, failure, and fission gas release behavior under the simulated RIA condition was studied in the tests. Fuel failure due to pellet-cladding mechanical interaction (PCMI) did not occur in the tests with typical domestic BWR fuel at burnups up to 56 GWd/tU, because they had limited cladding embrittlement due to hydrogen absorption of about 100 ppm or less. However, the cladding failure occurred in tests with fuel at a burnup of 61 GWd/tU, in which the peak hydrogen content in the cladding was above 150 ppm. This type of failure was observed for the first time in BWR fuels. The cladding failure occurred at fuel enthalpies of 260 to 360 J/g (62 to 86 cal/g), which were higher than the PCMI failure thresholds decided by the Japanese Nuclear Safety Commission. From post-test examinations of the failed fuel, it was found that the crack in the BWR cladding progressed in a manner different from the one in PWR cladding failed in earlier tests, owing to its more randomly oriented hydride distribution. Because of these differences, the BWR fuel was judged to have failed at hydrogen contents lower than those of the PWR fuel. Comparison of the test results with code calculations revealed that the PCMI failure was caused by thermal expansion of pellets, rather than by the fission gas expansion in the pellets. The gas expansion, however, was found to cause large cladding hoop deformation later after the cladding temperature escalated. (author)

Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

International Nuclear Information System (INIS)

Edwards, Robert; Huang, Zhengyu

2001-01-01

The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been expanded for boiling water reactor (BWR) out-of-phase behavior. During BWR out-of-phase oscillation half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. A description of the new HRS is given; three computers are employed to handle all the computations required, including real-time data processing and graph generation. BWR out-of-phase oscillation was successfully simulated. By adjusting the reactivity feedback gains from boiling channels to the TRIGA reactor and to the first harmonic mode power simulation, limit cycle can be generated with both reactor power and the simulated first harmonic power. A 3-D display of spatial power distributions of fundamental mode, first harmonic, and total powers over the reactor cross section is shown

Algoritmo evolucionário para otimização do plano de tratamento em radioterapia conformal 3D

Directory of Open Access Journals (Sweden)

Marco César Goldbarg

2009-08-01

Full Text Available O planejamento do tratamento por radioterapia tem por objetivo atingir um volume alvo com altas doses de radiação tomando cuidado para não expor órgãos sadios a doses elevadas. É, portanto, muito importante que se encontre um balanço ideal entre esses objetivos conflitantes. O presente trabalho relata um modelo de programação matemática multiobjetivo e introduz um Algoritmo Transgenético para o problema de seleção do direcionamento dos feixes de radiação no planejamento em radioterapia conformal 3D. A seleção das direções dos feixes é feita através de uma técnica denominada de isocentros variáveis. Com a finalidade de testar o potencial do algoritmo desenvolvido, realiza-se um experimento comparativo com um Algoritmo Genético Multiobjetivo. O experimento computacional obtém dados quantitativos e qualitativos que são analisados no trabalho.The radiotherapy treatment planning aims to achieve a target volume with high doses of radiation taking care not to expose healthy organs to high doses. It is therefore very important to find an optimal balance between these conflicting goals. This paper reports a mathematical model of multiobjective programming and presents a Transgenetic Algorithm for the problem of selecting the direction of radiation beams in 3D conformal radiotherapy planning. The selection of beams directions is done with a technique called variable isocenters. In order to test the potential of the developed algorithm, a comparative experiment with a multiobjective genetic algorithm was done. The computational experiment obtains quantitative and qualitative data that are analyzed in this paper.

Siemens Nuclear Power Corporation methods development for BWR/PWR reactor licensing

International Nuclear Information System (INIS)

Pruitt, D.W.

1992-01-01

This presentation addresses the Siemens Nuclear Power Corporation (SNP) perspective on the primary forces driving methods development in the nuclear industry. These forces are fuel design, computational environment and industry requirement evolution. The first segment of the discussion presents the SNP experience base. SNP develops, manufactures and licenses both BWR and PWR reload fuel. A review of this experience base highlights the accelerating rate at which new fuel designs are being introduced into the nuclear industry. The application of advanced BWR lattice geometries provides an example of fuel design trends. The second aspect of the presentation is the rapid evolution of the computing environment. The final subject in the presentation is the impact of industry requirements on code or methods development

Recent SCDAP/RELAP5 improvements for BWR severe accident simulations

International Nuclear Information System (INIS)

Griffin, F.P.

1995-01-01

A new model for the SCDAP/RELAP5 severe accident analysis code that represents the control blade and channel box structures in a boiling water reactor (BWR) has been under development since 1991. This model accounts for oxidation, melting, and relocation of these structures, including the effects of material interactions between B 4 C, stainless steel, and Zircaloy. This paper describes improvements that have been made to the BWR control blade/channel box model during 1994 and 1995. These improvements include new capabilities that represent the relocation of molten material in a more realistic manner and modifications that improve the usability of the code by reducing the frequency of code failures. This paper also describes a SCDAP/RELAP5 assessment calculation for the Browns Ferry Nuclear Plant design based upon a short-term station blackout accident sequence

Algoritmo genético para reducir el makespan en un flow shop híbrido flexible con máquinas paralelas no relacionadas y tiempos de alistamiento dependientes de la secuencia

Directory of Open Access Journals (Sweden)

Juan Camilo López-Vargas

2015-01-01

Full Text Available El artículo propone el algoritmo genético simple o estándar (AGS como enfoque de solución al problema de programación de produc - ción para un ambiente tipo flow shop híbrido flexible minimizando el makespan. La codificación del algoritmo propuesto permite obtener resultados con tiempos de cómputo bastante razonables y con un nivel de convergencia del makespan cercano al 2%, con mejores solu - ciones que un algoritmo alternativo diseñado para el mismo caso de programación de producción. A partir de los resultados obtenidos en el proceso de experimentación y del posterior análisis comparativo, se concluye que a partir del modelamiento más completo de las condiciones reales de producción, el algoritmo genético ejecuta la programación de producción reduciendo el tiempo máximo de procesamiento, o makespan. En futuros trabajos, el enfoque de investigación será la búsqueda de más escenarios alternativos de producción, con el fin de incrementar la aplicación de este tipo de herramientas y generar impacto en los entornos empresariales reales.

Logical model for the control of a BWR turbine;Modelo logico para el control de una turbina de un BWR

Energy Technology Data Exchange (ETDEWEB)

Vargas O, Y. [Universidad del Valle de Mexico, Campus Toluca, Av. Las Palmas No. 136, Col. San Jorge Pueblo Nuevo, 52140 Metepec, Estado de Mexico (Mexico); Amador G, R.; Ortiz V, J.; Castillo D, R., E-mail: yonaeton@hotmail.co [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2009-07-01

In this work a design of a logical model is presented for the turbine control of a nuclear power plant with a BWR like energy source. The model is sought to implement later on inside the thermal hydraulics code of better estimate RELAP/SCDAPSIM. The logical model is developed for the control and protection of the turbine, and the consequent protection to the BWR, considering that the turbine control will be been able to use for one or several turbines in series. The quality of the present design of the logical model of the turbine control is that it considers the most important parameters in the operation of a turbine, besides that they have incorporated to the logical model the secondary parameters that will be activated originally as true when the turbine model is substituted by a detailed model. The development of the logical model of a turbine will be of utility in the short and medium term to carry out analysis on the turbine operation with different operation conditions, of vapor extraction, specific steps of the turbine to feed other equipment s, in addition to analyze the separate and the integrated effect. (Author)

Signal analysis of acoustic and flow-induced vibrations of BWR main steam line

Energy Technology Data Exchange (ETDEWEB)

Espinosa-Paredes, G., E-mail: gepe@xanum.uam.mx [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Prieto-Guerrero, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Núñez-Carrera, A. [Comisión Nacional de Seguridad Nuclear y Salvaguardias, Doctor Barragán 779, Col. Narvarte, México, D.F. 03020 (Mexico); Vázquez-Rodríguez, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Centeno-Pérez, J. [Instituto Politécnico Nacional, Escuela Superior de Física y Matemáticas Unidad Profesional “Adolfo López Mateos”, Av. IPN, s/n, México, D.F. 07738 (Mexico); Espinosa-Martínez, E.-G. [Departamento de Sistemas Energéticos, Universidad Nacional Autónoma de México, México, D.F. 04510 (Mexico); and others

2016-05-15

Highlights: • Acoustic and flow-induced vibrations of BWR are analyzed. • BWR performance after extended power uprate is considered. • Effect of acoustic side branches (ASB) is analyzed. • The ASB represents a reduction in the acoustic loads to the steam dryer. • Methodology developed for simultaneous analyzing the signals in the MSL. - Abstract: The aim of this work is the signal analysis of acoustic waves due to phenomenon known as singing in Safety Relief Valves (SRV) of the main steam lines (MSL) in a typical BWR5. The acoustic resonance in SRV standpipes and fluctuating pressure is propagated from SRV to the dryer through the MSL. The signals are analyzed with a novel method based on the Multivariate Empirical Mode Decomposition (M-EMD). The M-EMD algorithm has the potential to find common oscillatory modes (IMF) within multivariate data. Based on this fact, we implement the M-EMD technique to find the oscillatory mode in BWR considering the measurements obtained collected by the strain gauges located around the MSL. These IMF, analyzed simultaneously in time, allow obtaining an estimation of the effects of the multiple-SRV in the MSL. Two scenarios are analyzed: the first is the signal obtained before the installation of the acoustic dampers (ASB), and the second, the signal obtained after installation. The results show the effectiveness of the ASB to damp the strong resonances when the steam flow increases, which represents an important reduction in the acoustic loads to the steam dryer.

Signal analysis of acoustic and flow-induced vibrations of BWR main steam line

International Nuclear Information System (INIS)

Espinosa-Paredes, G.; Prieto-Guerrero, A.; Núñez-Carrera, A.; Vázquez-Rodríguez, A.; Centeno-Pérez, J.; Espinosa-Martínez, E.-G.

2016-01-01

Highlights: • Acoustic and flow-induced vibrations of BWR are analyzed. • BWR performance after extended power uprate is considered. • Effect of acoustic side branches (ASB) is analyzed. • The ASB represents a reduction in the acoustic loads to the steam dryer. • Methodology developed for simultaneous analyzing the signals in the MSL. - Abstract: The aim of this work is the signal analysis of acoustic waves due to phenomenon known as singing in Safety Relief Valves (SRV) of the main steam lines (MSL) in a typical BWR5. The acoustic resonance in SRV standpipes and fluctuating pressure is propagated from SRV to the dryer through the MSL. The signals are analyzed with a novel method based on the Multivariate Empirical Mode Decomposition (M-EMD). The M-EMD algorithm has the potential to find common oscillatory modes (IMF) within multivariate data. Based on this fact, we implement the M-EMD technique to find the oscillatory mode in BWR considering the measurements obtained collected by the strain gauges located around the MSL. These IMF, analyzed simultaneously in time, allow obtaining an estimation of the effects of the multiple-SRV in the MSL. Two scenarios are analyzed: the first is the signal obtained before the installation of the acoustic dampers (ASB), and the second, the signal obtained after installation. The results show the effectiveness of the ASB to damp the strong resonances when the steam flow increases, which represents an important reduction in the acoustic loads to the steam dryer.

BWR Assembly Optimization for Minor Actinide Recycling

International Nuclear Information System (INIS)

Maldonado, G. Ivan; Christenson, John M.; Renier, J.P.; Marcille, T.F.; Casal, J.

2010-01-01

The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs). A top-level objective of the Advanced Fuel Cycle Systems Analysis program element of the DOE NERI program is to investigate spent fuel treatment and recycling options for current light water reactors (LWRs). Accordingly, this project targets to expand the traditional scope of nuclear fuel management optimization into the following two complementary specific objectives: (1) To develop a direct coupling between the pin-by-pin within-bundle loading control variables and core-wide (bundle-by-bundle) optimization objectives, (2) to extend the methodology developed to explicitly encompass control variables, objectives, and constraints designed to maximize minor actinide incineration in BWR bundles and cycles. The first specific objective is projected to 'uncover' dormant thermal margin made available by employing additional degrees of freedom within the optimization process, while the addition of minor actinides is expected to 'consume' some of the uncovered thermal margin. Therefore, a key underlying goal of this project is to effectively invest some of the uncovered thermal margin into achieving the primary objective.

Thermochemistry in BWR. An overview of applications of program codes and databases

International Nuclear Information System (INIS)

Hermansson, H-P.; Becker, R.

2010-01-01

The Swedish work on thermodynamics of metal-water systems relevant to BWR conditions has been ongoing since the 70ies, and at present time a compilation and adaptation of codes and thermodynamic databases are in progress. In the previous work, basic thermodynamic data were compiled for parts of the system Fe-Cr-Ni-Co-Zn-S-H 2 O at 25-300 °C. Since some thermodynamic information necessary for temperature extrapolations of data up to 300 °C was not published in the earlier works, these data have now been partially recalculated. This applies especially to the parameters of the HKF-model, which are used to extrapolate the thermodynamic data for ionic and neutral aqua species from 25 °C to BWR temperatures. Using the completed data, e.g. the change in standard Gibbs energy (ΔG 0 ) and the equilibrium constant (log K) can be calculated for further applications at BWR/LWR conditions. In addition a computer program is currently being developed at Studsvik for the calculation of equilibrium conductivity in high temperature water. The program is intended for PWR applications, but can also be applied to BWR environment. Data as described above will be added to the database of this program. It will be relatively easy to further develop the program e.g. to calculate Pourbaix diagrams, and these graphs could then be calculated at any temperature. This means that there will be no limitation to the temperatures and total concentrations (usually 10 -6 to 10 -8 mol/kg) as reported in earlier work. It is also easy to add a function generating ΔG 0 and log K values at selected temperatures. One of the fundamentals for this work was also to overview and collect publicly available thermodynamic program codes and databases of relevance for BWR conditions found in open sources. The focus has been on finding already done compilations and reviews, and some 40 codes and 15 databases were found. Codes and data-bases are often integrated and such a package is often developed for

Applied methods for mitigation of damage by stress corrosion in BWR type reactors; Metodos aplicados para la mitigacion del dano por corrosion bajo esfuerzo en reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Hernandez C, R.; Diaz S, A.; Gachuz M, M.; Arganis J, C. [Instituto Nacional de Investigaciones Nucleares, Gerencia de Ciencia de Materiales, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

1998-07-01

The Boiling Water nuclear Reactors (BWR) have presented stress corrosion problems, mainly in components and pipes of the primary system, provoking negative impacts in the performance of energy generator plants, as well as the increasing in the radiation exposure to personnel involucred. This problem has caused development of research programs, which are guided to find solution alternatives for the phenomena control. Among results of greater relevance the control for the reactor water chemistry stands out particularly in the impurities concentration and oxidation of radiolysis products; as well as the supervision in the materials selection and the stresses levels reduction. The present work presents the methods which can be applied to diminish the problems of stress corrosion in BWR reactors. (Author)

DESARROLLO DE UN SISTEMA DE NEURO-MERCADOTECNIA USANDO EL DISPOSITIVO EMOTIV-EPOC

OpenAIRE

Luz Ángela Moreno; César Augusto Peña; Oscar Eduardo Gualdron

2014-01-01

Este artículo presenta los avances concebidos en la realización de un sistema de neuro-mercadotecnia. El objetivo del sistema es permitir evaluar las emociones que presentan los televidentes al observar comerciales publicitarios. Se hace uso de la interfaz cerebro-computador Emotiv-EPOC para la adquisición de las neuro-señales. Se describen los algoritmos empleados para el análisis de las pautas publicitarias. Se presentan unas pruebas experimentales donde se comparan los resultados obtenidos...

Power plant design: ESBWR - the latest passive BWR

International Nuclear Information System (INIS)

Arnold, H.; Yadigaroglu, G.; Stoop, P.C.

1997-01-01

When General Electric said it would end development of its 670 MWe SBWR (Simplified Boiling Water Reactor), it was not quite the end of the story. Also on the drawing board at the time was the larger ESBWR (standing for either European or Economic Simplified BWR) whose goal was to provide the improved economic performance that the SBWR could not. (UK)

Análisis paramétrico de volúmenes arquitectónicos con algoritmos genéticos

Directory of Open Access Journals (Sweden)

Pedro Salcedo Lagos

2012-06-01

Full Text Available En el último tiempo el diseño arquitectónico ha venido evolucionando, debido entre otros aspectos, a la aparición de técnicas de diseño digital que permiten la generación de geometrías a partir de la definición de parámetros iniciales y la programación de las relaciones formales entre ellos. Los procesos de diseño basados en estas tecnologías permiten describir formas con capacidad de variar y adaptarse a requerimientos múltiples o a criterios de evaluación específicos, surgiendo así el problema de identificar la mejor solución arquitectónica, lo que se ha planteado en varias experiencias, gracias a la utilización de la técnica de algoritmos genéticos. En este trabajo se demuestra la posibilidad de implementar un análisis paramétrico de volúmenes arquitectónicos con algoritmos genéticos, logrando compatibilizar requerimientos funcionales, ambientales y estructurales, con un método efectivo de búsqueda para seleccionar una variedad de soluciones apropiadas gracias a las tecnologías digitales.

BWR fuel experience with zinc injection

International Nuclear Information System (INIS)

Levin, H.A.; Garcia, S.E.

1995-01-01

In 1982 a correlation between low primary recirculation system dose rates in BWR's and the presence of ionic zinc in reactor water was identified. The source of the zinc was primarily from Admiralty brass condensers. Plants with brass condensers are called ''natural zinc'' plants. Brass condensers were also a source of copper that was implicated in crude induced localized corrosion (CILC) fuel failures. In 1986 the first BWR intentionally injected zinc for the benefits of dose rate control. Although zinc alone was never implicated in fuel degradation of failures, a comprehensive fuel surveillance program was initiated to monitor fuel performance. Currently there are 14 plants that are injecting zinc. Six of these plants are also on hydrogen water chemistry. This paper describes the effect on both Zircaloy corrosion and the cruding characteristics as a result of these changes in water chemistry. Fuel rod corrosion was found to be independent of the specific water chemistry of the plants. The corrosion behavior was the same with the additions of zinc alone or zinc plus hydrogen and well within the operating experience for fuel without either of these additions. No change was observed in the amounts of crude deposited on the fuel rods, both for the adherent and loosely held deposits. One of the effects of the zinc addition was the trend to form more of the zinc rich iron spinel in the fuel deposits rather than the hematite deposits that are predominantly formed with non additive water chemistry

Programação funcional usando Java

OpenAIRE

Jorge Juan Zavaleta Gavidia

1997-01-01

Desde a introdução da World Wide Web para o mundo nos inícios de 1990, usando a Internet como uma rede para transferir dados, empregando uma forma de expressão chamada de Hipertexto, a qual liga as informações relacionadas e combinadas com multimídia, os Webs resultantes têm aberto novas possibilidades de expressão e comunicação. A quantidade de tráfego de dados na Web e o número de computadores ofertando informação vem crescendo dramaticamente, mas falta expressividade e qualidade interativa...

Assessment of boiling transition analysis code against data from NUPEC BWR full-size fine-mesh bundle tests

International Nuclear Information System (INIS)

Utsuno, Hideaki; Ishida, Naoyuki; Masuhara, Yasuhiro; Kasahara, Fumio

2004-01-01

Transient BT analysis code TCAPE based on mechanistic methods coupled with subchannel analysis has been developed for the evaluation on fuel integrity under abnormal operations in BWR. TCAPE consisted mainly of the drift-flux model, the cross-flow model, the film model and the heat transfer model. Assessment of TCAPE has been performed against data from BWR full-size fine-mesh bundle tests (BFBT), which consisted of two major parts: the void distribution measurement and the critical power measurement. Code and data comparison was made for void distributions with varying number of unheated rods in simulated actual fuel assembly. Prediction of steady-state critical power was compared with the measurement on full-scale bundle under a range of BWR operational conditions. Although the cross-sectional averaged void fraction was underestimated when it became lower, the accuracy was obtained that the averaged ratio 0.910 and its standard deviation 0.076. The prediction of steady-state critical power agreed well with the data in the range of BWR operations, where the prediction accuracy was obtained that the averaged ratio 0.997 and its standard deviation 0.043. These results demonstrated that TCAPE is well capable to predict two-phase flow distribution and liquid film dryout phenomena occurring in BWR rod bundles. Part of NUPEC BFBT database will be made available for an international benchmark exercise. The code assessment shall be continued against the OECD/NRC benchmark based on BFBT database. (author)

Dose rate reduction method for NMCA applied BWR plants

International Nuclear Information System (INIS)

Nagase, Makoto; Aizawa, Motohiro; Ito, Tsuyoshi; Hosokawa, Hideyuki; Varela, Juan; Caine, Thomas

2012-09-01

BRAC (BWR Radiation Assessment and Control) dose rate is used as an indicator of the incorporation of activated corrosion by products into BWR recirculation piping, which is known to be a significant contributor to dose rate received by workers during refueling outages. In order to reduce radiation exposure of the workers during the outage, it is desirable to keep BRAC dose rates as low as possible. After HWC was adopted to reduce IGSCC, a BRAC dose rate increase was observed in many plants. As a countermeasure to these rapid dose rate increases under HWC conditions, Zn injection was widely adopted in United States and Europe resulting in a reduction of BRAC dose rates. However, BRAC dose rates in several plants remain high, prompting the industry to continue to investigate methods to achieve further reductions. In recent years a large portion of the BWR fleet has adopted NMCA (NobleChem TM ) to enhance the hydrogen injection effect to suppress SCC. After NMCA, especially OLNC (On-Line NobleChem TM ), BRAC dose rates were observed to decrease. In some OLNC applied BWR plants this reduction was observed year after year to reach a new reduced equilibrium level. This dose rate reduction trends suggest the potential dose reduction might be obtained by the combination of Pt and Zn injection. So, laboratory experiments and in-plant tests were carried out to evaluate the effect of Pt and Zn on Co-60 deposition behaviour. Firstly, laboratory experiments were conducted to study the effect of noble metal deposition on Co deposition on stainless steel surfaces. Polished type 316 stainless steel coupons were prepared and some of them were OLNC treated in the test loop before the Co deposition test. Water chemistry conditions to simulate HWC were as follows: Dissolved oxygen, hydrogen and hydrogen peroxide were below 5 ppb, 100 ppb and 0 ppb (no addition), respectively. Zn was injected to target a concentration of 5 ppb. The test was conducted up to 1500 hours at 553 K. Test

CREACIÓN DEL ALGORITMO HEURÍSTICO PARA LA PLANIFICACIÓN Y PROGRAMACIÓN DE CARTERAS DE PROYECTOS CON RECURSOS LIMITADOS

Directory of Open Access Journals (Sweden)

Mª Antonia Simón Rodríguez

2014-12-01

Full Text Available Existen numerosas empresas que tienen implantada una Oficina de Proyectos y los gestionan de manera unificada. Uno de los principales problemas que se encuentran es la asignación de los recursos humanos a los proyectos de cara a cumplir con los plazos establecidos. El objetivo del presente artículo es crear un nuevo algoritmo heurístico que logre resolverlo en un entorno dinámico, que tenga en cuenta las incidencias que van sucediendo, buscando el cumplimiento de hitos intermedios, dando continuidad a los recursos en un mismo proyecto y asignando distintas prioridades. Se ha realizado un estudio de la literatura referente al entorno multiproyecto y para el caso particular de carteras de proyectos. El algoritmo resultante logra satisfacer los objetivos propuestos con gran eficacia y eficiencia, logrando cumplir el plazo para todos los proyectos con prioridad crítica, respetar las fechas intermedias y logrando la nivelación de los recursos humanos disponibles.

Vertical Drop of 44-BWR Waste Package With Lifting Collars

Energy Technology Data Exchange (ETDEWEB)

A.K. Scheider

2005-08-23

The objective of this calculation is to determine the structural response of a waste package (WP) dropped flat on its bottom from a specified height. The WP used for that purpose is the 44-Boiling Water Reactor (BWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. The Uncanistered Waste Disposal Container System is classified as Quality Level 1 (Ref. 4, page 7). Therefore, this calculation is subject to the requirements of the Quality Assurance Requirements and Description (Ref. 16). AP-3. 12Q, Design Calculations and Analyses (Ref. 11) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 44-BWR WP considered in this calculation and provides the potential dimensions and materials for that design.

Study on thermal performance and margins of BWR fuel elements

International Nuclear Information System (INIS)

Stosic, Zoran

1999-01-01

This paper contributes to developing a methodology of predicting and analyzing thermal performance and margins of Boiling Water Reactor (BWR) fuel assemblies under conditions of reaching high quality Boiling Crisis and subsequent post-dryout thermal hydraulics causing temperature excursion of fuel cladding. Operational margins against dryout and potential for increasing fuel performance with appropriate benefits are discussed. The philosophy of modeling with its special topics are demonstrated on the HECHAN (HEated CHannel ANalyzer) model as the state-of-art for thermal-hydraulics analysis of BWR fuel assemblies in pre- and post-dryout two-phase flow regimes. The scope of further work either being or has to be performed concerning implementation of new physical aspects, including domain extension of HECHAN model applications to the Pressurized Water Reactors (PWRs), is discussed. Finally, a comprehensive overview of the literature dealing with development of the model is given. (author)

Sophistication of operator training using BWR plant simulator

International Nuclear Information System (INIS)

Ohshiro, Nobuo; Endou, Hideaki; Fujita, Eimitsu; Miyakita, Kouji

1986-01-01

In Japanese nuclear power stations, owing to the improvement of fuel management, thorough maintenance and inspection, and the improvement of facilities, high capacity ratio has been attained. The thorough training of operators in nuclear power stations also contributes to it sufficiently. The BWR operator training center was established in 1971, and started the training of operators in April, 1974. As of the end of March, 1986, more than 1800 trainees completed training. At present, in the BWR operator training center, No.1 simulator of 800 MW class and No.2 simulator of 1100 MW class are operated for training. In this report, the method, by newly adopting it, good result was obtained, is described, that is, the method of introducing the feeling of being present on the spot into the place of training, and the new testing method introduced in retraining course. In the simulator training which is apt to place emphasis on a central control room, the method of stimulating trainees by playing the part of correspondence on the spot and heightening the training effect of multiple monitoring was tried, and the result was confirmed. The test of confirmation on the control board was added. (Kako, I.)

Control in fabrication of PWR and BWR type reactor fuel elements

International Nuclear Information System (INIS)

Gorskij, V.V.

1981-01-01

Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru

Development of a computerized operator support system for BWR power plant

International Nuclear Information System (INIS)

Monta, K.; Sekimizu, K.; Sato, N.; Araki, T.; Mori, N.

1985-01-01

A computerized operator support system for BWR power plant has been developed since 1980 supported by the Japanese government. The main functions of the systems are post trip operational guidance, disturbance analysis, standby system management, operational margin monitoring and control rod operational guidance. The former two functions aim at protection against incidents during operation of nuclear power plants and the latter three functions aim at their prevention. As the final stage of the development, these functions are combined with the plant supervision function and are organized as an advanced man-machine interface for BWR power plant. During the above process, operator task analyses are performed to enable synthesis of these support functions for right fit to operator tasks and to realize a hierarchical structure for CRT displays for right fit to operators cognitive needs. (author)

Prediction of droplet deposition around BWR fuel spacer by FEM flow analysis

International Nuclear Information System (INIS)

Yamamoto, Yasushi; Morooka, Shinichi

1997-01-01

The critical power of the BWR fuel assembly has been remarkably increased. That increase mainly depends on the improvement of the spacer which keeps fixed gaps between fuel rods. So far, these improvements have been carried out on the basis of what developers consider to be appropriate and the results of mockup tests of the BWR fuel assembly. However, continued reliance on these approaches for the development of a higher performance fuel assembly will prove time-consuming and costly. Therefore, it is hoped that the spacer effects for the critical power can be investigated by computer simulation, and it is significantly important to develop the critical power prediction method. Direct calculation of the two-phase flow in a BWR fuel channel s still difficult. Accordingly, a new method for predicting the critical power was proposed. Our method consists of CFD (computer fluid dynamics) code based on the single-phase flow analysis method and the subchannel analysis code. To verify our method, the critical power predictions for various spacer geometries were performed. The predicted results of the critical power were compared with the experimental data. The result of the comparison showed a good agreement and the applicability of our method for various spacer geometries. (author)

Water chemistry control and decontamination experience with TEPCO BWR`s and the measures planned for the future

Energy Technology Data Exchange (ETDEWEB)

Suzuki, N.; Miyamaru, K. [Tokyo Electric Power Co. (Japan)

1995-03-01

The new TEPCO BWR`s are capable of having the occupational radiation exposure controlled successfully at a low level by selecting low cobalt steel, using corrosion-resistant steel, employing dual condensate polishing systems, and controlling Ni/Fe ratio during operation. The occupational radiation exposure of the old BWR`s, on the other hand, remains high though reduced substantially through the use of low cobalt replacement steel and the partial addition of a filter in the condensate polishing system. Currently under review is the overall decontamination procedure for the old BWR`s to find out to measures needed to reduce the amount of crud that is and has been carried over into the nuclear reactor. The current status of decontamination is reported below.

METODOLOGÍA DE SOLUCIÓN PARA PLANEAMIENTO DE LA TRANSMISIÓN CONSIDERANDO INCERTIDUMBRE EN LA DEMANDA Y PROPUESTAS DE DIFERENTES CONDUCTORES

Directory of Open Access Journals (Sweden)

Andrés Hernando Domínguez Castaño

Full Text Available En este artículo se presenta una metodología de solución para resolver el problema de planeamiento estático de redes de transmisión de energía eléctrica, considerando incertidumbre en la demanda y selección de conductores en las líneas de transmisión que hacen parte de los nuevos corredores. Este problema de optimización se resuelve usando un algoritmo genético especializado que utiliza la lógica del algoritmo genético propuesto por Chu y Beasley, combinado con una técnica exacta. La metodología se prueba sobre el sistema eléctrico colombiano de 93 nodos y 155 líneas candidatas. Los resultados obtenidos mejoran la solución para el planeamiento estático del sistema eléctrico colombiano.

PREDICTIVE METHODS FOR STABILITY MARGIN IN BWR

OpenAIRE

MELARA SAN ROMÁN, JOSÉ

2016-01-01

[EN] Power and flow oscillations in a BWR are very undesirable. One of the major concerns is to ensure, during power oscillations, compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions ...

Valuation of power oscillations in a BWR after control rod banks withdrawal events

International Nuclear Information System (INIS)

Costa, A. L.; Pereira, C.; Da Silva, C. A. M.; Veloso, M. A. F.

2009-01-01

The out-of-phase mode of oscillation is a very challenging type of instability occurring in BWR (Boiling Water Reactor) and its study is relevant because of the safety implications related to the capability to promptly detect any such inadvertent occurrence by in-core neutron detectors, thus triggering the necessary countermeasures in terms of selected rod insertion or even reactor shutdown. In this work, control rod banks (CRB) withdrawal transient was considered to study the power instability occurring in a BWR. To simulate this transient, the control rod banks were continuously removed from the BWR core in different cases. The simulation resulted in a very large increase of power. To perform the instability simulations, the RELAP5/MOD3.3 thermal hydraulic system code was coupled with the PARCS/2.4 3D neutron kinetic code. Data from a real BWR, the Peach Bottom, have been used as reference conditions and reactor parameters. The trend of the mass flow rate, pressure, coolant temperature and the void fraction to four thermal hydraulic channels symmetrically located in the core with respect to the core centre, were taken. It appears that the velocity of the rod bank withdrawal is a very important aspect for reactor stability. The slowest CRB withdrawal (180 s) did not cause power perturbation while the fast removal (20 s) triggered a slow power oscillation that little by little amplified to reach levels of more 100% of the initial power after about 210 s. The investigation of the related thermo hydraulic parameters showed that the mass flow rate, the void fraction and also the coolant temperature began to oscillate at approximately the same time interval

Analysis of Core Physics Experiments on Irradiated BWR MOX Fuel in REBUS Program

International Nuclear Information System (INIS)

Yamamoto, Toru; Ando, Yoshihira; Hayashi, Yamato

2008-01-01

As part of analyses of experimental data of a critical core containing a irradiated BWR MOX test bundle in the REBUS program, depletion calculations was performed for the BWR MOX fuel assemblies from that the MOX test rods were selected by using a general purpose neutronics code system SRAC. The core analyses were carried out using SRAC and a continuous energy Monte Carlo code MVP. The calculated k eff s were compared with those of the core containing a fresh MOX fuel bundle in the program. The SRAC-diffusion calculation underestimates k eff s of the both cores by 1.0 to 1.3 %dk and the k eff s of MVP are 1.001. The difference in k eff between the irradiated BWR MOX test bundle core and the fresh MOX one is 0.4 %dk in the SRAC-diffusion calculation and 0.0 %dk in the MVP calculation. The calculated fission rate distributions are in good agreement with the measurement in the SRAC-diffusion and MVP calculations. The calculated neutron flux distributions are also in good agreement with the measurement. The calculated burnup reactivity in the both calculations well reproduce the measurements. (authors)

Prevention of organic iodide formation in BWR`s

Energy Technology Data Exchange (ETDEWEB)

Karjunen, T [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland); Laitinen, T; Piippo, J; Sirkiae, P [VTT Manufacturing Technology (Finland)

1996-12-01

During an accident, many different forms of iodine may emerge. Organic iodides, such as methyl iodide and ethyl iodide, are relatively volatile, and thus their appearance leads to increased concentration of gaseous iodine. Since organic iodides are also relatively immune to most accident mitigation measures, such as sprays and filters, they can affect the accident source term significantly even when only a small portion of iodine is in organic form. Formation of organic iodides may not be limited by the amount of organic substances available. Excessive amounts of methane can be produced, for example, during oxidation of boron carbide, which is used in BWR`s as a neutron absorber material. Another important source is cable insulation. In a BWR, a large quantity of cables is placed below the pressure vessel. Thus a large quantity of pyrolyse gases will be produced, should the vessel fail. Organic iodides can be formed as a result of many different reactions, but at least in certain conditions the main reaction takes place between an organic radical produced by radiolysis and elemental iodine. A necessary requirement for prevention of organic iodide production is therefore that the pH in the containment water pools is kept high enough to eliminate formation of elemental iodine. In a typical BWR the suppression pool water is usually unbuffered. As a result, the pH may be dominated by chemicals introduced during an accident. If no system for adding basic chemicals is operable, the main factor affecting pool water pH may be hydrochloric acid released during cable degradation. Should this occur, the conditions could be very favorable for production of elemental iodine and, consequently, formation of organic iodides. Although high pH is necessary for iodine retention, it could have also adverse effects. High pH may, for example, accelerate corrosion of containment materials and alter the characteristics of the solid corrosion products. (author) 6 figs., 1 tab., 13 refs.

OECD/NRC BWR Turbine Trip Transient Benchmark as a Basis for Comprehensive Qualification and Studying Best-Estimate Coupled Codes

International Nuclear Information System (INIS)

Ivanov, Kostadin; Olson, Andy; Sartori, Enrico

2004-01-01

An Organisation for Economic Co-operation and Development (OECD)/U.S. Nuclear Regulatory Commission (NRC)-sponsored coupled-code benchmark has been initiated for a boiling water reactor (BWR) turbine trip (TT) transient. Turbine trip transients in a BWR are pressurization events in which the coupling between core space-dependent neutronic phenomena and system dynamics plays an important role. In addition, the available real plant experimental data make this benchmark problem very valuable. Over the course of defining and coordinating the BWR TT benchmark, a systematic approach has been established to validate best-estimate coupled codes. This approach employs a multilevel methodology that not only allows for a consistent and comprehensive validation process but also contributes to the study of different numerical and computational aspects of coupled best-estimate simulations. This paper provides an overview of the OECD/NRC BWR TT benchmark activities with emphasis on the discussion of the numerical and computational aspects of the benchmark

Requests on domestic nuclear data library from BWR design

International Nuclear Information System (INIS)

Maruyama, Hiromi

2003-01-01

Requests on the domestic nuclear data library JENDL and activities of the Nuclear Data Center have been presented from the perspective of BWR design and design code development. The requests include a standard multi-group cross section library, technical supports, and clarification of advantage of JENDL as well as requests from physical aspects. (author)

AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

Energy Technology Data Exchange (ETDEWEB)

Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

2016-05-15

Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

Uso de redes neuronales artificiales para mejorar el tránsito en las vías

Directory of Open Access Journals (Sweden)

Edwin Mejía

2016-11-01

Full Text Available Este artículo pretende dar una idea de cómo las Redes Neuronales Artificiales (ANNs, una técnica de la Inteligencia Artificial (IA, se puede acoplar a resolver el problema del tráfico en las vías de la ciudad de Riobamba, provincia de Chimborazo, país Ecuador, usando cuatro semáforos estándar, a través de un ejemplo se indica el uso de componentes electrónicos como FPGAs (Field Programmable Gate array y los sensores en este campo, a través de detectar y contar autos puede dar mayor fluidez al tráfico. La vía que mayor cantidad de autos tenga se dará mayor prioridad para la luz verde. Usando el algoritmo de aprendizaje de un Perceptrón Simple.

A study of heat capacity temperature limit of BWR

International Nuclear Information System (INIS)

Wang, Shih-Jen; Chen, Jyh-Jun; Chien, Chun-Sheng; Teng, Jyh-Tong

2012-01-01

Highlights: ► The purpose of this study is to verify the HCTL. ► MAAP4 was used as code to generate a realistic and convenient HCTL. ► The current HCTL curve causes confusing in reading data. ► The revised HCTL curves developed in this study. ► Users can obtain important parameters from the revised HCTL without confusion and interpolation. - Abstract: Heat capacity temperature limit (HCTL) is an important parameter for operation of BWR. Current version of the HCTL was derived, based on simple model of computation aids (CA) of BWR owners’ group (BWROG). However, some parts of the current HCTL are confusing to the users in reading data. The purpose of this study is to verify the HCTL by applying the MAAP4 code to the field of emergency operating procedure (EOP). The trends of HCTL generated by MAAP4 code are consistent with those obtained from CA. A series of revised HCTL evaluated at various times after scram are provided and the confusing part is eliminated.

System control model of a turbine for a BWR

International Nuclear Information System (INIS)

Vargas O, Y.; Amador G, R.; Ortiz V, J.; Castillo D, R.; Delfin L, A.

2009-10-01

In this work is presented a design of a control system of a turbine for a nuclear power plant with a BWR like energy source. The model seeks to implement later on at thermal hydraulics code of better estimate RELAP/SCDAPSIM. The model is developed for control and protection of turbine, and the consequent protection to the BWR, considering that the turbine control could be employed for one or several turbines in series. The quality of present designs of control pattern of turbine it is that it considers the parameters more important in the operation of a turbine besides that is has incorporated at control the secondary parameters that will be activated originally as true when the turbine model is substituted by a model more detailed. The development of control model of a turbine will be good in short and medium term to realize analysis about the operation of turbine with different operation conditions, of vapor extraction specific steps of turbine to feed other equipment s, besides analyzing the separate effect and integrated effect. (Author)

A study of heat capacity temperature limit of BWR

Energy Technology Data Exchange (ETDEWEB)

Wang, Shih-Jen, E-mail: sjenwang@iner.gov.tw [Institute of Nuclear Energy Research (INER), 1000, Wunhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Chen, Jyh-Jun [Department of Mechanical Engineering, Chung Yuan Christian University, 200, Chung Pei Rd., Chung Li City, Taoyuan County 32023, Taiwan (China); Chien, Chun-Sheng [Institute of Nuclear Energy Research (INER), 1000, Wunhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Teng, Jyh-Tong [Department of Mechanical Engineering, Chung Yuan Christian University, 200, Chung Pei Rd., Chung Li City, Taoyuan County 32023, Taiwan (China)

2012-02-15

Highlights: Black-Right-Pointing-Pointer The purpose of this study is to verify the HCTL. Black-Right-Pointing-Pointer MAAP4 was used as code to generate a realistic and convenient HCTL. Black-Right-Pointing-Pointer The current HCTL curve causes confusing in reading data. Black-Right-Pointing-Pointer The revised HCTL curves developed in this study. Black-Right-Pointing-Pointer Users can obtain important parameters from the revised HCTL without confusion and interpolation. - Abstract: Heat capacity temperature limit (HCTL) is an important parameter for operation of BWR. Current version of the HCTL was derived, based on simple model of computation aids (CA) of BWR owners' group (BWROG). However, some parts of the current HCTL are confusing to the users in reading data. The purpose of this study is to verify the HCTL by applying the MAAP4 code to the field of emergency operating procedure (EOP). The trends of HCTL generated by MAAP4 code are consistent with those obtained from CA. A series of revised HCTL evaluated at various times after scram are provided and the confusing part is eliminated.

TRAB, a transient analysis program for BWR. Part 1

International Nuclear Information System (INIS)

Rajamaeki, Markku.

1980-03-01

TRAB is a transient analysis program for BWR. The present report describes its principles. The program has been developed from TRAWA-program. It models the interior of the pressure vessel and related subsystems of BWR viz. reactor core, recirculation loop including the upper part of the vessel, recirculation pumps, incoming and outgoing flow systems, and control and protection systems. Concerning core phenomena and all flow channel hydraulics the submodels are one-dimensional of main features. The geometry is very flexible. The program has been made particularly to simulate various reactivity transients, but it is applicable more generally to reactor incidents and accidents in which no flow reversal or no emptying of the circuit must occur below the water level. The program is extensively supplied by input and output capabilities. The user can act upon the simulation of a transient by defining external disturbances, scheduled timevariations for any system variable, by modeling new subsystems, which are representable with ordinary linear differential equations, and by defining relations of functional form between system variables. The run of the program can be saved and restarted. (author)

Impact of advanced BWR core physics method on BWR core monitoring

Energy Technology Data Exchange (ETDEWEB)

Moon, H; Wells, A [Siemens Power Corporation, Richland (United States)

2000-07-01

Siemens Power Corporation recently initiated development of POWERPLEX{sup TM}-III for delivery to the Grand Gulf Nuclear Power Station. The main change introduced in POWERPLEX{sup TM}-III as compared to its predecessor POWERPLEX{sup TM}-II is the incorporation of the advances BWR core simulator MICROBURN-B2. A number of issues were identified and evaluated relating to the implementation of MICROBURN-B2 and its impact on core monitoring. MICROBURN-B2 demands about three to five times more memory and two to three times more computing time than its predecessor MICROBURN-B in POWERPLEX {sup TM}-II. POWERPLEX{sup TM}-III will improve thermal margin prediction accuracy and provide more accurate plant operating conditions to operators than POWERPLEX{sup TM}-II due to its improved accuracy in predicted TIP values and critical k-effective. The most significant advantage of POWERPLEX{sup TM}-III is its capability to monitor a relaxed rod sequence exchange operation. (authors)

El método de los elementos finitos para el modelado de la ecuación de ondas con un procesador vectorial

OpenAIRE

Seron, F.; Sanz, F.; Kindelan, M.; C.Perez, C.Perez

1990-01-01

El objetivo de este trabajo es analizar los aspectos computacionales del Método de los Elementos Finitos para la resolución de las ecuaciones de onda elásticas. Se analizan las técnicas numéricas necesarias desde el punto de vista de la precisión, prestaciones y necesidades de almacenamiento cuando se implementan en procesadores escalares y vectoriales con gran capacidad de almacenamiento. El método se ha implementado en un IBM 3090 con procesador vectorial usando diferentes algoritmos para l...

Development of membrane moisture separator for BWR off-gas system

International Nuclear Information System (INIS)

Ogata, H.; Kawamura, S.; Kumasaka, M.; Nishikubo, M.

2001-01-01

In BWR plant off-gas treatment systems, dehumidifiers are used to maintain noble gas adsorption efficiency in the first half of the charcoal hold-up units. From the perspective of simplifying and reducing the cost of such a dehumidification system, Japanese BWR utilities and plant fabricators have been developing a dehumidification system employing moisture separation membrane of the type already proven in fields such as medical instrumentation and precision measuring apparatus. The first part of this development involved laboratory testing to simulate the conditions found in an actual off-gas system, the results of which demonstrated satisfactory results in terms of moisture separation capability and membrane durability, and suggested favorable prospects for application in actual off-gas systems. Further, in-plant testing to verify moisture separation capability and membrane durability in the presence of actual gases is currently underway, with results so far suggesting that the system is capable of obtaining good moisture separation capability. (author)

Flux and power distributions in BWR multi-bundle fuel arrays

International Nuclear Information System (INIS)

Cheng, H.S.

1976-02-01

Multi-bundle calculations have been performed in order to shed some light on an abnormal TIP trace recently discovered in a BWR/3. Transport theory was employed to perform the calculations with ENDF/B-IV data. The results indicate that a strong variation of the TIP reading does exist along the narrow water gap of a BWR due to the steep gradient of the thermal neutron flux; the maxima occurring at the intersections of the water gaps and the minima in between. Using this characteristic behavior of the TIP reading, together with the observed normal TIP trace, the abnormal behavior of the affected TIP trace exhibiting three peaks along the channel was roughly simulated. The calculations confirmed that the observed TIP trace anomaly was caused by the severe bending of the affected instrument tube as was actually discovered. The effect of hot water intrusion into the TIP guide tube, as well as that of loading the new 8 x 8 reload bundles, was also evaluated

Damage by radiation in structural materials of BWR reactor vessels

International Nuclear Information System (INIS)

Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

2002-01-01

The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA Mark III Salazar reactor and separately with Ni +3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A 2 ). (Author)

Studies of fragileness in steels of vessels of BWR reactors

International Nuclear Information System (INIS)

Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

2003-01-01

The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA MARK lll reactor and separately with Ni +3 ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A 2 . (Author)

Accident sequence analysis for a BWR [Boiling Water Reactor] during low power and shutdown operations

International Nuclear Information System (INIS)

Whitehead, D.W.; Hake, T.M.

1990-01-01

Most previous Probabilistic Risk Assessments have excluded consideration of accidents initiated in low power and shutdown modes of operation. A study of the risk associated with operation in low power and shutdown is being performed at Sandia National Laboratories for a US Boiling Water Reactor (BWR). This paper describes the proposed methodology for the analysis of the risk associated with the operation of a BWR during low power and shutdown modes and presents preliminary information resulting from the application of the methodology. 2 refs., 2 tabs

Monitorização de transplante cardíaco usando análises do eletrograma intracavitário

Directory of Open Access Journals (Sweden)

BROFMAN Paulo Roberto S.

1997-01-01

Full Text Available Uma série de registros do eletrograma intracavitário tem sido utilizada para monitorização não invasiva da rejeição em pacientes transplantados usando um marcapasso de dupla câmara e eletrodos endocavitários revestidos com estrutura fractal. Os sinais têm sido avaliados usando o sistema CHARM (Computerized Heart Acute Rejection Monitoring _ sistema computadorizado para monitorização da rejeição cardíaca aguda. Os relatórios obtidos com este sistema contêm curvas com parâmetros sensíveis à rejeição, que demonstram uma boa correlação com a clínica e os resultados das biópsias convencionais. A monitorização a longo prazo, usando estas análises, mostrou ser uma ferramenta valiosa no acompanhamento destes pacientes.

Seismic PRA of a BWR plant

International Nuclear Information System (INIS)

Nishio, Masahide; Fujimoto, Haruo

2014-01-01

Since the occurrence of nuclear power plant accidents in the Fukushima Daichi nuclear power station, the regulatory framework on severe accident (SA) has been discussed in Japan. The basic concept is to typify and identify the accident sequences leading to core/primary containment vessel (PCV) damage and to implement SA measures covering internal and external events extensively. As Japan is an earthquake-prone country and earthquakes and tsunami are important natural external events for nuclear safety of nuclear power plants, JNES performed the seismic probabilistic risk assessment (PRA) on a typical nuclear power plant and evaluated the dominant accident sequences leading to core/PCV damage to discuss dominant scenarios of severe accident (SA). The analytical models and the results of level-1 seismic PRA on a 1,100 MWe BWR-5 plant are shown here. Seismic PRA was performed for a typical BWR5 plant. Initiating events with large contribution to core damage frequency are the loss of all AC powers (station blackout) and the large LOCA. The top of dominant accident sequences is the simultaneous occurrence of station blackout and large LOCA. Important components to core damage frequency are electric power supply equipment. It needs to keep in mind that the results are influenced on site geologic characteristic to a greater or lesser. In the process of analysis, issues such as conservative assumptions related to damages of building or structure and success criteria for excessive LOCA are left to be resolved. These issues will be further studied including thermal hydric analysis in the future. (authors)

Experimental study on reduced moderation BWR with Advanced Recycle System (BARS)

International Nuclear Information System (INIS)

Hiraiwa, K.; Yoshioka, K.; Yamamoto, Y.; Akiba, M.; Yamaoka, M.; Abe, N.; Mimatsu, J.

2004-01-01

Experimental study has been done for reduced-moderation spectrum boiling water reactor named BARS (BWR with Advanced Recycle System). The critical assembly experiment for triangular tight uranium lattice has been done in TOSHIBA critical assembly (NCA). Experimental method based on modified conversion ratio was adopted to evaluate the void reactivity effect. Void fraction was simulated by formed polystyrene in this experiment. The measured void coefficient for tight uranium lattice agreed with calculation. The thermal hydraulic test study has been done to study the coolability of BARS lattice. Visual test and high-pressure thermal hydraulic test have been done as the thermal hydraulic test. Visual test has indicated the flow behavior for BARS lattice is same as that of current BWR. The high-pressure thermal hydraulic test has indicated the applicability of modified Arai's correlation to the BARS lattice. (authors)

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

DESARROLLO DE UN SISTEMA DE NEURO-MERCADOTECNIA USANDO EL DISPOSITIVO EMOTIV-EPOC

Directory of Open Access Journals (Sweden)

Luz Ángela Moreno

2014-11-01

Full Text Available Este artículo presenta los avances concebidos en la realización de un sistema de neuro-mercadotecnia. El objetivo del sistema es permitir evaluar las emociones que presentan los televidentes al observar comerciales publicitarios. Se hace uso de la interfaz cerebro-computador Emotiv-EPOC para la adquisición de las neuro-señales. Se describen los algoritmos empleados para el análisis de las pautas publicitarias. Se presentan unas pruebas experimentales donde se comparan los resultados obtenidos por métodos tradicionales con las propuestas.

Unificando los criterios de sepsis neonatal tardía: propuesta de un algoritmo de vigilancia diagnóstica

Science.gov (United States)

Zea-Vera, Alonso; Turin, Christie G.; Ochoa, Theresa J.

2015-01-01

Las infecciones constituyen una de las principales causas de muerte en el periodo neonatal. El diagnóstico de sepsis neonatal representa un gran desafío ya que los recién nacidos presentan signos clínicos muy inespecíficos y los exámenes auxiliares tienen una baja sensibilidad. Con el objetivo de mejorar el diagnóstico correcto de esta patología proponemos un algoritmo de vigilancia diagnóstica para sepsis neonatal tardía en el Perú y países de la región. El algoritmo permite clasificar a los episodios como sepsis confirmada, probable o posible, y sobretodo busca identificar aquellos episodios que no corresponden a sepsis, evitando calificar otras patologías como “sepsis”. Un mejor diagnóstico permitiría tener tasas más reales de sepsis neonatal, mejorar el uso de antibióticos y evitar sus efectos negativos en el recién nacido, así como una visión más exacta de su impacto en la salud pública. PMID:25123879

Organización del conocimiento de la asignatura “estructuras de datos y algoritmosâ€� para ingeniería informática basada en mapas conceptuales

OpenAIRE

SOLER PELLICER, YOLANDA; LEZCANO BRITO, MATEO GERÓNIMO

2009-01-01

Las asignaturas de Programación y Algoritmos deben lograr que los estudiantes escriban algoritmos eficientes. Ese objetivo se ve afectado por el hecho de que los estudiantes no saben escoger y diseñar las estructuras de datos adecuadas para representar la información. Tomando en cuenta este hecho, el Centro de Estudios de Informática Educativa (C EIE ) de la Universidad de Granma y el Centro de Estudios Informáticos (CEI) de la Universidad Central “Marta Abreuâ€� de Las Villas, Cuba, propon...

Results of the Simulator smart against synthetic signals using a model of reduced order of BWR with additive and multiplicative noise; Resultados del simulador smart frente a senales sinteticas utilizando un modelo de orden reducido de BWR con ruido aditivo y multiplicativo

Energy Technology Data Exchange (ETDEWEB)

Munoz-Cobo, J. L.; Montesino, M. E.; Pena, J.; Escriva, A.; Melara, J.

2011-07-01

Results of SMART-simulator front of synthetic signals with models of reduced order of BWR with additive and multiplicative noise Under the SMART project, which aims to monitor the signals Cofrentes nuclear plant, we have developed a signal generator of synthetics BWR that will allow together real signals of plant the validation of the monitor.

Aging assessment of BWR control rod drive systems

International Nuclear Information System (INIS)

Greene, R.H.

1992-01-01

This Phase 1 Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assesses the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of NPRDS failure cases attributed to the CRD system, and (4) personal information exchange. As part of this study, nearly 3,500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation has been conducted that summarizes the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented regarding specific actions that utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities

Aging assessment of BWR control rod drive systems

International Nuclear Information System (INIS)

Greene, R.H.

1991-01-01

This study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assess the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the CRD system, and (4) personal information exchange with industry experts. As part of this study, nearly 3500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation was conducted to summarize the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented that identify specific actions utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain maintenance practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities. 5 refs., 8 figs., 2 tabs

Appraisal of BWR plutonium burners for energy centers

International Nuclear Information System (INIS)

Williamson, H.E.

1976-01-01

The design of BWR cores with plutonium loadings beyond the self-generation recycle (SGR) level is investigated with regard to their possible role as plutonium burners in a nuclear energy center. Alternative plutonium burner approaches are also examined including the substitution of thorium for uranium as fertile material in the BWR and the use of a high-temperature gas reactor (HTGR) as a plutonium burner. Effects on core design, fuel cycle facility requirements, economics, and actinide residues are considered. Differences in net fissile material consumption among the various plutonium-burning systems examined were small in comparison to uncertainties in HTGR, thorium cycle, and high plutonium-loaded LWR technology. Variation in the actinide content of high-level wastes is not likely to be a significant factor in determining the feasibility of alternate systems of plutonium utilization. It was found that after 10,000 years the toxicity of actinide high-level wastes from the plutonium-burning fuel cycles was less than would have existed if the processed natural ores had not been used for nuclear fuel. The implications of plutonium burning and possible future fuel cycle options on uranium resource conservation are examined in the framework of current ERDA estimates of minable uranium resources

Aging assessment of BWR control rod drive systems

International Nuclear Information System (INIS)

Greene, R.H.

1991-01-01

This Phase 1 Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assesses the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of NPRDS failure cases attributed to the CRD system, and (4) personal information exchange. As part of this study, nearly 3,500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation has been conducted that summarizes the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented regarding specific actions that utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities

Safety evaluation of BWR off-gas treatment systems

International Nuclear Information System (INIS)

Schultz, R.J.; Schmitt, R.C.

1975-01-01

Some of the results of a safety evaluation performed on current generic types of BWR off-gas treatment systems including cooled and ambient temperature adsorber beds and cryogenics are presented. The evaluation covered the four generic types of off-gas systems and the systems of five major vendors. This study was part of original work performed under AEC contract for the Directorate of Regulatory Standards. The analysis techniques employed for the safety evaluation of these systems include: Fault Tree Analysis; FMECA (Failure Mode Effects and Criticality Analysis); general system comparisons, contaminant, system control, and design adequacy evaluations; and resultant Off-Site Dose Calculations. The salient areas presented are some of the potential problem areas, the approach that industry has taken to mitigate or design against potential upset conditions, and areas where possible deficiencies still exist. Potential problem areas discussed include hydrogen detonation, hydrogen release to equipment areas, operator/automatic control interface, and needed engineering evaluation to insure safe system operation. Of the systems reviewed, most were in the category of advanced or improved over that commonly in use today, and a conclusion from the study was that these systems offer excellent potential for noble gas control for BWR power plants where more stringent controls may be specified -- now or in the future. (U.S.)

A detailed BWR recirculation loop model for RELAP

Energy Technology Data Exchange (ETDEWEB)

Araiza-Martínez, Enrique, E-mail: enrique.araiza@inin.gob.mx; Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.mx; Castillo-Durán, Rogelio, E-mail: rogelio.castillo@inin.gob.mx

2017-01-15

Highlights: • A new detailed BWR recirculation loop model was developed for RELAP. • All jet pumps, risers, manifold, suction and control valves, and recirculation pump are modeled. • Model is tested against data from partial blockage of two jet pumps. • For practical applications, simulation results showed good agreement with available data. - Abstract: A new detailed geometric model of the whole recirculation loop of a BWR has been developed for the code RELAP. This detailed model includes the 10 jet pumps, 5 risers, manifold, suction and control valves, and the recirculation pump, per recirculation loop. The model is tested against data from an event of partial blockage at the entrance nozzle of one jet pump in both recirculation loops. For practical applications, simulation results showed good agreement with data. Then, values of parameters considered as figure of merit (reactor power, dome pressure, core flow, among others) for this event are compared against those from the common 1 jet pump per loop model. The results show that new detailed model led to a closer prediction of the reported power change. The detailed recirculation loop model can provide more reliable boundary condition data to a CFD models for studies of, for example, flow induced vibration, wear, and crack initiation.

Potenciando el aprendizaje proactivo con ILIAS&WebQuest: aprendiendo a paralelizar algoritmos con GPUs

OpenAIRE

Santamaría, J.; Espinilla, M.; Rivera, A. J.; Romero, S.

2010-01-01

Arquitectura de Computadores es una asignatura troncal de segundo ciclo de la titulación de Ingeniería de Telecomunicación (P.E. 2004) de la Universidad de Jaén, que desde el curso académico 2009/10 cuenta con una metodología de aprendizaje proactivo para motivar al alumno en la realización de las prácticas. En concreto, se ha abordado la enseñanza de la materia de paralelización de algoritmos haciendo uso de GPUs de tarjetas gráficas convencionales. Además, se ha d...

Algoritmos relax-and-cut para problemas de programação inteira 0-1

OpenAIRE

Victor Fernandes Cavalcante

2008-01-01

Resumo: Uma das principais motivações para o estudo de Otimização Discreta reside no elevado número de problemas do nosso cotidiano representáveis através de modelos de Otimização Inteira e Combinatória. Em particular, muitos destes problemas podem ser formulados com Programação Inteira 0-1, o que desperta especial interesse em técnicas capazes de resolver tais modelos. Dentre as inúmeras formas de solução atualmente disponíveis para problemas desta natureza, os algoritmos baseados na técnica...

Séries temporais de NDVI do sensor SPOT Vegetation e algoritmo SAM aplicados ao mapeamento de cana‑de‑açúcar

Directory of Open Access Journals (Sweden)

Luiz Eduardo Vicente

2012-09-01

Full Text Available O objetivo deste trabalho foi avaliar o mapeamento de área de cana‑de‑açúcar por meio de série temporal, de seis anos de dados do índice de vegetação por diferença normalizada (NDVI, oriundos do sensor Vegetation, a bordo do satélite "système pour l'observation de la Terre" (SPOT. Três classes de cobertura do solo (cana‑de‑açúcar, pasto e floresta, do Estado de São Paulo, foram selecionadas como assinaturas espectro‑temporais de referência, que serviram como membros extremos ("endmembers" para classificação com o algoritmo "spectral angle mapper" (SAM. A partir desta classificação, o mapeamento da área de cana‑de‑açúcar foi realizado com uso de limiares na imagem-regra do SAM, gerados a partir dos valores dos espectros de referência. Os resultados mostram que o algoritmo SAM pode ser aplicado a séries de dados multitemporais de resolução moderada, o que permite eficiente mapeamento de alvo agrícola em escala mesorregional. Dados oficiais de áreas de cana‑de‑açúcar, para as microrregiões paulistas, apresentam boa correlação (r² = 0,8 com os dados obtidos pelo método avaliado. A aplicação do algoritmo SAM mostrou ser útil em análises temporais. As séries temporais de NDVI do sensor SPOT Vegetation podem ser utilizadas para mapeamento da área de cana‑de‑açúcar em baixa resolução.

Maintenance of BWR control rod drive mechanisms

International Nuclear Information System (INIS)

Greene, R.H.

1991-01-01

Control rod drive mechanism (CRDM) replacement and rebuilding is one of the highest dose, most physically demanding, and complicated maintenance activities routinely accomplished by BWR utilities. A recent industry workshop sponsored by the Oak Ridge National Laboratory, which dealt with the effects of CRDM aging, revealed enhancements in maintenance techniques and tooling which have reduced ALARA, improved worker comfort and productivity, and have provided revised guidelines for CRDM changeout selection. Highlights of this workshop and ongoing research on CRDM aging are presented in this paper

Sensitiaztion of austenitic stainless steels and its significance as regards stress-corrosion cracking of BWR pipe systems

International Nuclear Information System (INIS)

Roberts, W.; Otterberg, R.

1984-05-01

A critical literature evaluation dealing with sensitization of austenitic stainless steels and its importance in the context of intergranular stress-corrosion cracking (IGSCC) in high-temperature, oxygenated water is presented. The factors influencing the degree of sensitization are discussed, principally for type-304 stainless steels, both as regards sensitization arising as a result of isothermal holding within the critical temperature range and weld sensitization. The phenomenon of low-temperature sensitization is described and its potential significance under BWR operating conditions speculated upon. The principal features of and mechanisms controlling IGSCC of sensitized 304 steels in BWR-type environments are reviewed and some thoughts are given to the relevance of laboratory SCC testing in predicting the occurrence of cracking in actual BWR systems. Finally various countermeasures against IGSCC in existing and projected reactors are presented and discussed. (Author)

Um sistema para o ensino e aprendizagem de algoritmos utilizando um companheiro de aprendizagem colaborativo

OpenAIRE

Petry, Patrícia Gerent

2005-01-01

Dissertação (mestrado) - Universidade Federal de Santa Catarina, Centro Tecnológico. Programa de Pós-graduação em Ciência da Computação Esta dissertação apresenta um modelo computacional de um sistema de suporte ao ensino e aprendizagem no domínio de algoritmos. O sistema inclui um companheiro de aprendizagem virtual, que utiliza a Modelagem Baseada em Restrições (Constraint-Based Modelling) como forma de representação do conhecimento e raciocínio. Os Sistemas Companheiro de Aprendizagem (...

Selección y evaluación de algoritmos para clasificación de documentos

OpenAIRE

Vergara García, Paula María

2014-01-01

Este trabajo ha consistido en un estudio sobre la clasificación de documentos en castellano, basándose en el estándar IPTC. Para ello se ha utilizado un conjunto de ejemplos formado por documentos previamente etiquetados. Partiendo de los distintos tipos de aprendizaje, los algoritmos de clasificación, las herramientas de las que se disponía para la clasificación y las características de conjunto de ejemplos, se decidió utilizar las librerías proporcionadas por las herramientas Mahout, Had...

Algoritmo distribuido de compresión de datos : Una experiencia comparativa con sockets y PVM

OpenAIRE

Russo, Claudia Cecilia; Ramón, Hugo Dionisio; Anderson, Alfredo; Dirazar, Delio; De Giusti, Armando Eduardo

1996-01-01

Es notoria la importancia actual de la compresión de datos en al transmisión de información en redes, especialmente información multimedial Aspectos como optimización del uso del canal, tiempos de repuesta, seguridad y reducción de tráfico, justifican las tareas de investigación y desarrollo en este tema, así como al evolución de los recursos tecnológicos disponibles para implementar nuevas técnicas. En este contexto, se presentan resultados experimentales obtenidos de un algoritmo de compres...

Algoritmos de visión para la estimación robusta de pose 3D

OpenAIRE

Iglesias García, Marcos

2010-01-01

El desarrollo de nuevos algoritmos en el campo de la visión artificial permite dotar a los robots móviles de una mayor autonomía en su funcionamiento. Si el autómata conoce el espacio que le rodea, es capaz de interactuar con el medio y realizar múltiples tareas. Tradicionalmente, los robots están equipados con microprocesadores, ordenadores a bordo que les permiten procesar la información que reciben y actuar en consecuencia. Sin embargo, carecen de inteligencia, lo que les impide reaccionar...

Application of eddy current inspection to the Inconel weld of BWR internals

International Nuclear Information System (INIS)

Machida, Eiji; Yusa, Noritaka

2004-01-01

In order to definite the basic specifications of application of ECT (Eddy Current Test) to Inconel weld of BWR internals, the inspection and numerical analysis were carried out. The characteristics of the existing ECT probe were studied by making sample as same as CRD stud tube, measuring the relative permeability and electric conductivity of Inconel and alloy and evaluating ECT probe. On the basis of the results obtained, the basic specifications were determined and a new eddy current probe for inspection was designed and produced. The new ECT probe was able to detect small notch in Inconel weld, to classify the defects by eddy current inspection signal and sizing the length and depth. It is concluded that the new ECT probe is able to apply the Inconel weld of BWR internals. (S.Y.)

IFPE/IFA-432, Fission Gas Release, Mechanical Interaction BWR Fuel Rods, Halden

International Nuclear Information System (INIS)

Turnbull, J.A.

1996-01-01

Description: It contains data from experiments that have been performed at the IFE/OECD Halden Reactor Project, available for use in fuel performance studies. It covers experiments on thermal performance, fission product release, clad properties and pellet clad mechanical interaction. It includes also experimental data relevant to high burn-up behaviour. IFA-432: Measurements of fuel temperature response, fission gas release and mechanical interaction on BWR-type fuel rods up to high burn-ups. The assembly featured several variations in rod design parameters, including fuel type, fuel/cladding gap size, fill gas composition (He and Xe) and fuel stability. It contained 6 BWR-type fuel rods with fuel centre thermocouples at two horizontal planes, rods were also equipped with pressure transducers and cladding extensometers. Only data from 6 rods are compiled here

Feasibility study on development of plate-type heat exchanger for BWR plants

International Nuclear Information System (INIS)

Ohyama, Nobuhiro; Suda, Kenichi; Ogata, Hiroshi; Matsuda, Shinichi; Nagasaka, Kazuhiro; Fujii, Toshi; Nozawa, Toshiya; Ishihama, Kiyoshi; Higuchi, Tomokazu

2004-01-01

In order to apply plate-type heat exchanger to RCW, TCW and FPC system in BWR plants, heat test and seismic test of RCW system heat exchanger sample were carried out. The results of these tests showed new design plate-type heat exchanger satisfied the fixed pressure resistance and seismic resistance and keep the function. The evaluation method of seismic design was constructed and confirmed by the results of tests. As anti-adhesion measure of marine organism, an ozone-water circulation method, chemical-feed method and combination of circulation of hot water and air bubbling are useful in place of the chlorine feeding method. Application of the plate-type heat exchanger to BWR plant is confirmed by these investigations. The basic principles, structure, characteristics, application limit and reliability are stated. (S.Y.)

TVA experience in BWR reload design and licensing

International Nuclear Information System (INIS)

Robertson, J.D.

1986-01-01

TVA has developed and implemented the capability to perform BWR reload core design and licensing analyses. The advantages accruing from this capability include the tangible cost-savings from performing reload analyses in-house. Also, ''intangible'' benefits such as increased operating flexibility and the ability to accommodate multivendor fuel designs have been demonstrated. The major disadvantage with performing in-house analyses is the cost associated with development and maintenance of the analytical methods and staff expertise

Aggressive chemical decontamination tests on small valves from the Garigliano BWR

International Nuclear Information System (INIS)

Bregani, F.

1990-01-01

In order to check the effectiveness of direct chemical decontamination on small and complex components, usually considered for storage without decontamination because of the small amount, some tests were performed on the DECO experimental loop. Four small stainless steel valves from the primary system of the Garigliano BWR were decontaminated using mainly aggressive chemicals such as HC1, HF, HNO 3 and their mixtures. On two valves, before the treatment with aggressive chemicals, a step with soft chemical (oxalic and citric acid mixture) was performed in order to see whether a softening action enhances the following aggressive decontamination. Moreover, in order to increase as much as possible the decontamination effectiveness, a decontamination process using ultrasounds jointly with aggressive chemicals was investigated. After an intensive laboratory testing programme, two smaller stainless steel valves from the primary system of the Garigliano BWR were decontaminated using ultrasounds in aggressive chemical solutions

Analysis of void reactivity measurements in full MOX BWR physics experiments

International Nuclear Information System (INIS)

Ando, Yoshihira; Yamamoto, Toru; Umano, Takuya

2008-01-01

In the full MOX BWR physics experiments, FUBILA, four 9x9 test assemblies simulating BWR full MOX assemblies were located in the center of the core. Changing the in-channel moderator condition of the four assemblies from 0% void to 40% and 70% void mock-up, void reactivity was measured using Amplified Source Method (ASM) technique in the subcritical cores, in which three fission chambers were located. ASM correction factors necessary to express the consistency of the detector efficiency between measured core configurations were calculated using collision probability cell calculation and 3D-transport core calculation with the nuclear data library, JENDL-3.3. Measured reactivity worth with ASM correction factor was compared with the calculated results obtained through a diffusion, transport and continuous energy Monte Carlo calculation respectively. It was confirmed that the measured void reactivity worth was reproduced well by calculations. (author)

Status report: Intergranular stress corrosion cracking of BWR core shrouds and other internal components

International Nuclear Information System (INIS)

1996-03-01

On July 25, 1994, the US Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 94-03 to obtain information needed to assess compliance with regulatory requirements regarding the structural integrity of core shrouds in domestic boiling water reactors (BWRs). This report begins with a brief description of the safety significance of intergranular stress corrosion cracking (IGSCC) as it relates to the design and function of BWR core shrouds and other internal components. It then presents a brief history of shroud cracking events both in the US and abroad, followed by an indepth summary of the industry actions to address the issue of IGSCC in BWR core shrouds and other internal components. This report summarizes the staff's basis for issuing GL 94-03, as well as the staff's assessment of plant-specific responses to GL 94-03. The staff is continually evaluating the licensee inspection programs and the results from examinations of BWR core shrouds and other internal components. This report is representative of submittals to and evaluations by the staff as of September 30, 1995. An update of this report will be issued at a later date

Analysis CFD for the hydrogen transport in the primary containment of a BWR; Analisis CFD para el transporte de hidrogeno en la contencion primaria de un reactor BWR

Energy Technology Data Exchange (ETDEWEB)

Jimenez P, D. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Gomez T, A. M., E-mail: guerreroazteca_69@hotmail.com [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2014-10-15

This study presents a qualitative and quantitative comparison among the CFD GASFLOW and OpenFOAM codes which are related with the phenomenon of hydrogen transport and other gases in the primary containment of a Boiling Water Reactor (BWR). GASFLOW is a commercial license code that is well validated and that was developed in Germany for the analysis of the gases transport in containments of nuclear reactors. On the other hand, OpenFOAM is an open source code that offers several evaluation solvers for different types of phenomena; in this case, the solver reacting-Foam is used to analyze the hydrogen transport inside the primary containment of the BWR. The results that offer the solver reacting-Foam of OpenFOAM are evaluated in the hydrogen transport calculation and the results are compared with those of the program of commercial license GASFLOW to see if is viable the use of the open source code in the case of the hydrogen transport in the primary containment of a BWR. Of the obtained results so much quantitative as qualitative some differences were identified between both codes, the differences (with a percentage of maximum error of 4%) in the quantitative results are small and they are considered acceptable for this analysis type, also, these differences are attributed mainly to the used transport models, considering that OpenFOAM uses a homogeneous model and GASFLOW uses a heterogeneous model. (Author)

Effect of the inlet throttling on the thermal-hydraulic instability of the natural circulation BWR

International Nuclear Information System (INIS)

Furuya, Masahiro; Inada, Fumio; Yoneda, Kimitoshi

1997-01-01

Although it is well-established that inlet restriction has a stabilizing for forced circulation BWR, the effect of inlet on the thermal-hydraulic stability of natural circulation BWR remains unknown since increasing inlet restriction affect thermal-hydraulic stability due to reduction of the recirculation flow rate. Therefore experiments have been conducted to investigate the effect of inlet restriction on the thermal-hydraulic stability. A test facility used in this experiments was designed and constructed to have non-dimensional values which are nearly equal to those of natural circulation BWR. Experimental results showed that driving force of the natural circulation was described as a function of heat flux and inlet subcooling independent of inlet restriction. Stability maps in reference to the channel inlet subcooling, heat flux were presented for various inlet restriction which were carried out by an analysis based on the homogeneous flow various using this function. Instability region during the inlet subcooling shifted to the higher inlet subcooling with increasing inlet restriction and became larger with increasing heat flux. (author)

Development of a coordinated control system for BWR nuclear power plant and HVDC transmission system

International Nuclear Information System (INIS)

Ishikawa, M.; Hara, T.; Hirayama, K.; Sekiya, K.

1986-01-01

The combined use of dc and ac transmissions or so-called hybrid transmission was under study, employing both dc and ac systems to enable stable transmission of 10,000 MW of electric power generated by the BWR nuclear plant, scheduled to be built about 800 km away from the center of the load. It was thus necessary to develop a hybrid power transmission control system, the hybrid power transmission system consisting of a high voltage dc transmission system (HVDC) and an ultrahigh ac transmission system (UHVAC). It was also necessary to develop a control system for HVDC transmission which protects the BWR nuclear power plant from being influenced by any change in transmission mode that occurs as a result of faults on the UHVAC side when the entire power of the BWR plant is being sent by the HVDC transmission. This paper clarifies the requirements for the HVDC system control during hybrid transmission and also during dc transmission. The control method that satisfies these requirements was studied to develop a control algorithm

CECP, Decommissioning Costs for PWR and BWR

International Nuclear Information System (INIS)

Bierschbach, M.C.

1997-01-01

1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

A BWR licensing experience in the USA

International Nuclear Information System (INIS)

Powers, J.; Ogura, C.; Arai, K.; Thomas, S.; Mookhoek, B.

2015-09-01

The US-Advanced Boiling Water Reactor (A BWR), certified by the United States Nuclear Regulatory Commission (US NRC), is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3-4 project has finished the US NRC technical review of the Cola through the final meeting of the Advisory Committee on Reactor Safeguards (ACRS), and the Final Safety Evaluation Report (FSER) is scheduled to be issued by the US NRC in the middle of 2015. The next steps are to support the Mandatory Hearing process, and voting by the NRC commissioners on the motion to grant the Combined License, which is scheduled beginning of 2016 according to US NRC schedule as of March 30, 2015. This paper summarizes the history and progress of the US-A BWR licensing, including the experiences of the Licensee, Nina, and Toshiba as the Epc team worked through the Code of Federal Regulations Title 10 (10-Cfr) Part 52 process, and provides some perspectives on how the related licensing material would also be of value within a 10-Cfr Part 50, two-step process to minimize schedule and financial risks which could arise from ongoing technical developments and regulatory reviews. (Author)

A BWR licensing experience in the USA

Energy Technology Data Exchange (ETDEWEB)

Powers, J.; Ogura, C. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Arai, K. [Toshiba Corporation, Yokohama, Kanagawa (Japan); Thomas, S.; Mookhoek, B., E-mail: jim.powers@toshiba.com [Nuclear Innovation North America, Lake Jackson, Texas (United States)

2015-09-15

The US-Advanced Boiling Water Reactor (A BWR), certified by the United States Nuclear Regulatory Commission (US NRC), is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3-4 project has finished the US NRC technical review of the Cola through the final meeting of the Advisory Committee on Reactor Safeguards (ACRS), and the Final Safety Evaluation Report (FSER) is scheduled to be issued by the US NRC in the middle of 2015. The next steps are to support the Mandatory Hearing process, and voting by the NRC commissioners on the motion to grant the Combined License, which is scheduled beginning of 2016 according to US NRC schedule as of March 30, 2015. This paper summarizes the history and progress of the US-A BWR licensing, including the experiences of the Licensee, Nina, and Toshiba as the Epc team worked through the Code of Federal Regulations Title 10 (10-Cfr) Part 52 process, and provides some perspectives on how the related licensing material would also be of value within a 10-Cfr Part 50, two-step process to minimize schedule and financial risks which could arise from ongoing technical developments and regulatory reviews. (Author)

BWR type reactors

International Nuclear Information System (INIS)

Watanabe, Shoichi

1986-01-01

Purpose: To enable to remove water not by way of mechanical operation in a reactor core and improve the fuel economy in BWR type reactors. Constitution: A hollow water removing rod of a cross-like profile made of material having a smaller neutron absorption cross section than the moderator is disposed to the water gap for each of unit structures composed of four fuel assemblies, and water is charged and discharged to and from the water removing rod. Water is removed from the water removing rod to decrease the moderators in the water gap to carry out neutron spectrum shift operation from the initial to the medium stage of reactor core cycles. At the final stage of the cycle, airs in the water removing rod are extracted and the moderator is introduced. The moderator is filled and the criticality is maintained with the accumulated nuclear fission materials. The neutron spectrum shift operation can be attained by eliminating hydrothermodynamic instability and using a water removing rod of a simple structure. (Horiuchi, T.)

Transmutation of minor actinide using BWR fueled mixed oxide

International Nuclear Information System (INIS)

Susilo, Jati

2000-01-01

Nuclear spent fuel recycle has a strategic importance in the aspect of nuclear fuel economy and prevention of its spread-out. One among other application of recycle is to produce mixed oxide fuel (Mo) namely mixed Plutonium and uranium oxide. As for decreasing the burden of nuclear high level waste (HLW) treatment, transmutation of minor actinide (MA) that has very long half life will be carried out by conversion technique in nuclear reactor. The purpose of this study was to know influence of transition fuel cell regarding the percent weight of transmutation MA in the BWR fueled MOX. Calculation of cell BWR was used SRAC computer code, with assume that the reactor in equilibrium. The percent weight of transmutation MA to be optimum by increasing the discharge burn-up of nuclear fuel, raising ratio of moderator to fuel volume (Vm/Vf), and loading MA with percent weight about 3%-6% and also reducing amount of percent weight Pu in MOX fuel. For mixed fuel standard reactor, reactivity value were obtained between about -50pcm ∼ -230pcm for void coefficient and -1.8pcm ∼ -2.6pcm for fuel temperature coefficient

Evaluation of thermal margin during BWR neutron flux oscillation

International Nuclear Information System (INIS)

Takeuchi, Yutaka; Takigawa, Yukio; Chuman, Kazuto; Ebata, Shigeo

1992-01-01

Fuel integrity is very important, from the view point of nuclear power plant safety. Recently, neutron flux oscillations were observed at several BWR plants. The present paper describes the evaluations of the thermal margin during BWR neutron flux oscillations, using a three-dimensional transient code. The thermal margin is evaluated as MCPR (minimum critical power ratio). The LaSalle-2 event was simulated and the MCPR during the event was evaluated. It was a core-wide oscillation, at which a large neutron flux oscillation amplitude was observed. The results indicate that the MCPR had a sufficient margin with regard to the design limit. A regional oscillation mode, which is different from a core-wide oscillation, was simulated and the MCPR response was compared with that for the LaSalle-2 event. The MCPR decrement is greater in the regional oscillation, than in the core wide -oscillation, because of the sensitivity difference in a flow-to-power gain. A study was carried out about regional oscillation detectability, from the MCPR response view point. Even in a hypothetically severe case, the regional oscillation is detectable by LPRM signals. (author)

Logical model for the control of a BWR turbine

International Nuclear Information System (INIS)

Vargas O, Y.; Amador G, R.; Ortiz V, J.; Castillo D, R.

2009-01-01

In this work a design of a logical model is presented for the turbine control of a nuclear power plant with a BWR like energy source. The model is sought to implement later on inside the thermal hydraulics code of better estimate RELAP/SCDAPSIM. The logical model is developed for the control and protection of the turbine, and the consequent protection to the BWR, considering that the turbine control will be been able to use for one or several turbines in series. The quality of the present design of the logical model of the turbine control is that it considers the most important parameters in the operation of a turbine, besides that they have incorporated to the logical model the secondary parameters that will be activated originally as true when the turbine model is substituted by a detailed model. The development of the logical model of a turbine will be of utility in the short and medium term to carry out analysis on the turbine operation with different operation conditions, of vapor extraction, specific steps of the turbine to feed other equipment s, in addition to analyze the separate and the integrated effect. (Author)

Normalización de imágenes de placas vehiculares a través de corrección geométrica

OpenAIRE

Choez Álvarez, Carlos Leonardo; Salas Guerrero, Steve Fernando; Vintimilla, Boris X.

2013-01-01

El proyecto en general consiste en implementar un sistema de control de acceso vehicular mediante el reconocimiento del número de placa de manera automática usando una cámara y algoritmos de procesamiento de imágenes incluyendo Reconocimiento Óptico de Caracteres (OCR). Para el reconocimiento de las placas se tomara una imagen al vehículo al momento de entrar a un parqueadero específico. Este proyecto consta de cinco partes: 1. Detección y extracción de placas. 2. Normalización de imágenes...

Uma metodologia para classificação de dados nominais baseada no processo KDD

OpenAIRE

Souza, Rodrigo Clemente Thom de

2013-01-01

Resumo: A classificação de padrões é um problema de aprendizado supervisionado do campo da ciência conhecido como Reconhecimento de Padrões (RP), através do qual se deseja discriminar instâncias de dados em diferentes classes. A solução para este problema é obtida por meio de algoritmos (classificadores) que buscam por padrões de relacionamento entre classes em casos conhecidos (treinamento), usando tais relações para classificar casos desconhecidos (teste). O desempenho em termos de acurácia...

Geração de Isolinhas, com dados obtidos por levantamento GPS/L1L2, mediante a técnica de Redes Neurais Artificiais - DOI: 10.4025/actascitechnol.v28i2.1195

Directory of Open Access Journals (Sweden)

Elaine Cristine Barros de Souza

2006-03-01

Full Text Available O objetivo deste trabalho é mostrar uma alternativa para o processo de interpolação empregado em Modelagem Digital do Terreno. Utilizando-se dados de levantamentos GPS coletados por meio de posicionamento relativo cinemático realizou-se um processo de interpolação por meio de Redes Neurais Artificiais (RNA. Os resultados foram comparados com os dados obtidos usando-se o algoritmo de interpolação Inverso do Quadrado de uma Distância (IQD para avaliar o comportamento das grades geradas por meio de análises qualitativas (isolinhas e MDT e quantitativas (resíduos. Conclui-se que o método testado é viável se comparado com o algoritmo IQD e, portanto, o uso de RNA apresenta-se adequado para a interpolação de dados GPS provenientes do processamento de fase das portadoras L1 e L2

Power ramp tests of BWR-MOX fuels

International Nuclear Information System (INIS)

Asahi, K.; Oguma, M.; Higuchi, S.; Kamimua, K.; Shirai, Y.; Bodart, S.; Mertens, L.

1996-01-01

Power ramp test of BWR-MOX and UO 2 fuel rods base irradiated up to about 60 GWd/t in Dodewaard reactor have been conducted in BR2 reactor in the framework of the international DOMO programme. The MOX pellets were provided by BN (MIMAS process) and PNC (MH method). The MOX fuel rods with Zr-liner and non-liner cladding and the UO 2 fuel rods with Zr-liner cladding remained intact during the stepwise power ramp tests to about 600 W/cm, even at about 60 GWd/t

Algoritmos de métodos numéricos codificados en lenguaje de programación C++ para la solución de problemas de análisis numérico

OpenAIRE

León Zárate, Elmer Alberto

2014-01-01

La presente investigación tuvo como propósito la elaboración de un sistema computacional en el lenguaje de programación C++ que permita codificar los algoritmos para la solución de problemas de análisis numéricos. Este trabajo de investigación tiene por título: "ALGORITMOS DE MÉTODOS NUMÉRICOS CODIFICADOS EN LENGUAJE DE PROGRAMACION C++ PARA LA SOLUCION DE PROBLEMAS DE ANÁLISIS NUMÉRICO" y ha sido preparado para apoyar la formación de los estudiantes de ciencias e ingeniería, quienes podrán a...

Developing and modeling of the 'Laguna Verde' BWR CRDA benchmark

International Nuclear Information System (INIS)

Solis-Rodarte, J.; Fu, H.; Ivanov, K.N.; Matsui, Y.; Hotta, A.

2002-01-01

Reactivity initiated accidents (RIA) and design basis transients are one of the most important aspects related to nuclear power reactor safety. These events are re-evaluated whenever core alterations (modifications) are made as part of the nuclear safety analysis performed to a new design. These modifications usually include, but are not limited to, power upgrades, longer cycles, new fuel assembly and control rod designs, etc. The results obtained are compared with pre-established bounding analysis values to see if the new core design fulfills the requirements of safety constraints imposed on the design. The control rod drop accident (CRDA) is the design basis transient for the reactivity events of BWR technology. The CRDA is a very localized event depending on the control rod insertion position and the fuel assemblies surrounding the control rod falling from the core. A numerical benchmark was developed based on the CRDA RIA design basis accident to further asses the performance of coupled 3D neutron kinetics/thermal-hydraulics codes. The CRDA in a BWR is a mostly neutronic driven event. This benchmark is based on a real operating nuclear power plant - unit 1 of the Laguna Verde (LV1) nuclear power plant (NPP). The definition of the benchmark is presented briefly together with the benchmark specifications. Some of the cross-sections were modified in order to make the maximum control rod worth greater than one dollar. The transient is initiated at steady-state by dropping the control rod with maximum worth at full speed. The 'Laguna Verde' (LV1) BWR CRDA transient benchmark is calculated using two coupled codes: TRAC-BF1/NEM and TRAC-BF1/ENTREE. Neutron kinetics and thermal hydraulics models were developed for both codes. Comparison of the obtained results is presented along with some discussion of the sensitivity of results to some modeling assumptions

Recycling systems for BWR type reactors

International Nuclear Information System (INIS)

Takagi, Akio; Yamamoto, Fumiaki; Fukumoto, Ryuji.

1986-01-01

Purpose: To stabilize the coolant flowing characteristics and reactor core reactivity. Constitution: The recycling system in a BWR type reactor comprises a recycling pump disposed to the outside of a reactor pressure vessel, a ring header connected to the recycling pump through main pipe ways, and a plurality of pipes branched from and connected with the ring header and connected to a plurality of jet pumps within the pressure vessel. Then, by making the diameter for the pipeways of each of the branched pipes different from each other, the effective cross-sectional area is varied to thereby average the coolant flow rate supplied to each of the jet pumps. (Seki, T.)

MINIMIZACIÓN DE UNA FUNCIÓN DE ORDEN P MEDIANTE UN ALGORITMO GENÉTICO // MINIMIZING A FUNCTION OF ORDER P USING A GENETIC ALGORITHM

Directory of Open Access Journals (Sweden)

Rómulo Castillo Cárdenas

2013-06-01

Full Text Available In this work we consider the problem OVO (order value optimization. The problem we address is to minimize f with x 2 by a genetic algorithm that by its very nature has the advantage over existing continuous optimization methods, to nd global minimizers. We illustrate the application of this algorithm on examples considered showing its e ectiveness in solving them.// RESUMEN En el presente trabajo consideramos el problema OVO (order value optimization. El problema que abordamos consiste entonces en minimizar f con x 2 por medio de un algoritmo gen etico que por su naturaleza intrínseca tiene la ventaja, sobre métodos de optimización continua existentes, de encontrar minimizadores globales. Ilus- tramos la aplicación de este algoritmo sobre ejemplos considerados mostrando su eficacia en la resolución de los mismos.

Evaluation of PWR and BWR pin cell benchmark results

International Nuclear Information System (INIS)

Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J.; Hoogenboom, J.E.; Leege, P.F.A. de; Voet, J. van der; Verhagen, F.C.M.

1991-12-01

Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs

Evaluation of PWR and BWR pin cell benchmark results

Energy Technology Data Exchange (ETDEWEB)

Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

1991-12-01

Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs.

Evaluation of PWR and BWR pin cell benchmark results

Energy Technology Data Exchange (ETDEWEB)

Pilgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

1991-12-01

Benchmark results of the Dutch PINK working group on the PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs.; 9 figs.; 30 tabs.

Applied methods for mitigation of damage by stress corrosion in BWR type reactors

International Nuclear Information System (INIS)

Hernandez C, R.; Diaz S, A.; Gachuz M, M.; Arganis J, C.

1998-01-01

The Boiling Water nuclear Reactors (BWR) have presented stress corrosion problems, mainly in components and pipes of the primary system, provoking negative impacts in the performance of energy generator plants, as well as the increasing in the radiation exposure to personnel involucred. This problem has caused development of research programs, which are guided to find solution alternatives for the phenomena control. Among results of greater relevance the control for the reactor water chemistry stands out particularly in the impurities concentration and oxidation of radiolysis products; as well as the supervision in the materials selection and the stresses levels reduction. The present work presents the methods which can be applied to diminish the problems of stress corrosion in BWR reactors. (Author)

BWR stability using a reducing dynamical model; Estabilidad de un BWR con un modelo dinamico reducido

Energy Technology Data Exchange (ETDEWEB)

Ballestrin Bolea, J M; Blazquez Martinez, J B

1990-07-01

BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical structure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations is non-linear. Simple parametric calculation of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author)

Fission product model for BWR analysis with improved accuracy in high burnup

International Nuclear Information System (INIS)

Ikehara, Tadashi; Yamamoto, Munenari; Ando, Yoshihira

1998-01-01

A new fission product (FP) chain model has been studied to be used in a BWR lattice calculation. In attempting to establish the model, two requirements, i.e. the accuracy in predicting burnup reactivity and the easiness in practical application, are simultaneously considered. The resultant FP model consists of 81 explicit FP nuclides and two lumped pseudo nuclides having the absorption cross sections independent of burnup history and fuel composition. For the verification, extensive numerical tests covering over a wide range of operational conditions and fuel compositions have been carried out. The results indicate that the estimated errors in burnup reactivity are within 0.1%Δk for exposures up to 100GWd/t. It is concluded that the present model can offer a high degree of accuracy for FP representation in BWR lattice calculation. (author)

European BWR R and D cluster for innovative passive safety systems

International Nuclear Information System (INIS)

Hicken, E.F.; Lensa, W. von

1996-01-01

The main technological innovation trends for future nuclear power plants tend towards a broader use of passive safety systems for the prevention, mitigation and managing of severe accident scenarios. Several approaches have been undertaken in a number of European countries to study and demonstrate the feasibility and charateristics of innovative passive safety systems. The European BWR R and D Cluster combines those experimental and analytical efforts that are mainly directed to the introduction of passive safety systems into boiling water reactor technology. The Cluster is grouped around thermohydraulic test facilities in Europe for the qualification of innovative BWR safety systems, also taking into account especially the operating experience of the nuclear power plant Dodewaard and other BWRs, which already incorporated some passive safety features. The background, the objectives, the structure of the project and the work programme are presented in this paper as well as an outline of the significance of the expected results. (orig.) [de

Thermal-hydraulics stability of natural circulation BWR under startup. Flashing effects

International Nuclear Information System (INIS)

Hu, Rui; Kazimi, Mujid S.

2009-01-01

To help achieve the necessary natural circulation flow, a fairly long chimney is installed in a boiling natural circulation reactor like the ESBWR. In such systems, thermal-hydraulic stability during low pressure start-up should be examined while considering the flashing induced by the pressure drop in the channel and the chimney due to gravity head. In this work, a BWR stability analysis code in the frequency domain, named FISTAB (Flashing-Induced STability Analysis for BWR), was developed to address the issue of flashing-induced instability. A thermal-hydraulics non-homogeneous equilibrium model (NHEM) based on a drift flux formulation along with a lumped fuel dynamics model is incorporated in the work. The vapor generation rate is derived from the mixture energy conservation equation while considering the effect of flashing. The functionality of the FISTAB code was confirmed by comparison to experimental results from SIRIUS-N facility at CRIEPI, Japan. Both stationary and perturbation results agree well with the experimental results. (author)

ALGORITMO DE BÚSQUEDA TABÚ ESPECIALIZADO APLICADO AL DISEÑO DE REDES SECUNDARIAS DE ENERGÍA ELÉCTRICA

Directory of Open Access Journals (Sweden)

Víctor Mario Vélez Marín

Full Text Available En este artículo se presenta una metodología para solucionar el problema del planeamiento de sistemas de distribución secundarios empleando como técnica de solución el algoritmo de Búsqueda Tabú. El problema se formula como un modelo no lineal entero-mixto, en el cual se tienen en cuenta la ubicación y capacidad de nuevos elementos (transformadores de distribución y tramos de red primaria y secundaria, reubicación de transformadores de distribución existentes, aumento de capacidades de elementos existentes, reconfiguración de red secundaria y balance de fases. Adicionalmente, se consideran los costos asociados a la conexión entre red primaria y secundaria y las pérdidas de energía en transformadores. Se emplean dos casos de prueba; en el primero se realizan ensayos comparativos con el algoritmo genético de Chu-Beasley para verificar la eficiencia del método propuesto y, en el segundo, se analizan los resultados obtenidos en un sistema de distribución colombiano. En ambos casos los resultados obtenidos son de gran calidad, lo que respalda lo propuesto en este trabajo.

Algoritmo de programação de máquinas individuais com penalidades distintas de adiantamento e atraso

Directory of Open Access Journals (Sweden)

Emerson C. Colin

2000-06-01

Full Text Available Neste trabalho consideramos o problema de máquina única, com datas de entrega e penalidades de adiantamento e atraso distintas para cada ordem. Considerando que a seqüência seja predefinida, o objetivo a ser alcançado é a minimização da soma das diferenças (adiantamentos ou atrasos penalizadas das ordens. Este trabalho é apresentado como uma generalização do algoritmo de programação de Garey et al. (1988. Através de uma estrutura computacional denominada fila de prioridade, este novo algoritmo permite a elaboração de um programa em tempo O(nlogn, enquanto que o melhor encontrado na literatura atualmente é de tempo O(n².In this work we consider the one machine problem, with distinct due-dates and penalties for earliness and tardiness. For a previously defined sequence, we utilize the sum of weighted lateness (earliness or tardiness as objective function. This work is presented as a generalization of the Garey et al. (1988 scheduling algorithm. Using a computational structure called heap, this algorithm allows a schedule construction in O(nlogn time while the best found in literature runs in O(n².

Detection of failed fuel rods in shrouded BWR fuel assemblies

International Nuclear Information System (INIS)

Baero, G.; Boehm, W.; Goor, B.; Donnelly, T.

1988-01-01

A manipulator and an ultrasonic testing (UT) technique were developed to identify defective fuel rods in shrouded BWR fuel assemblies. The manipulator drives a UT probe axially through the bottom tie plate into the water channels between the fuel rods. The rotating UT probe locates defective fuel rods by ingressed water which attenuates the UT-signal. (author)

Assessment of severe accident prevention and mitigation features: BWR, Mark II containment design

International Nuclear Information System (INIS)

Lehner, J.R.; Hsu, C.J.; Eltawila, F.; Perkins, K.R.; Luckas, W.J.; Fitzpatrick, R.G.; Pratt, W.T.

1988-07-01

Plant features and operator actions, which have been found to be important in either preventing or mitigating severe accidents in BWRs with Mark II containments (BWR Mark II's) have been identified. These features and actions were developed from insights derived from reviews of in-depth risk assessments performed specifically for the Limerick and Shoreham plants and from other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the BWR Mark II to severe-accident containment loads were also noted. In addition, those features of a BWR Mark II, which are important for preventing core damage and are available for mitigating fission-product release to the environment were also identified. This report is issued to provide focus to an analyst examining an individual plant. This report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Mark II plants. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance

Examination of minor actinide annihilation by BWR core

International Nuclear Information System (INIS)

Hida, Kazuki

1995-01-01

From the viewpoint of reducing burden for disposing high level waste generated from spent fuel, the examination of recycling minor actinide (MA) to reactors and reducing its accumulation has been advanced. In this study, the possibility of annihilation in the case of recycling it to a BWR was examined. The main MAs are 237 Np, 241 Am, 243 Am, 242 Cm, and 244 Cm. However, as for Cm isotopes, the half life is short, the amount of generation is small, and the rate of neutron emission is high, therefore, those are disposed as waste, and 237 Np, 241 Am and 243 Am were taken as the objects of recycling. In order to grasp the basic characteristics in the case of recycling MAs to a BWR, MAs were added to UO 2 fuel, MOX fuel and HCR fuel and burned, and the nuclear conversion characteristics were examined. As the result, it was found that they were converted to short half life nuclides, and as the neutron spectra were softer, the rate of annihilation was higher. In the case of recycling MAs by concentrating to a specific reactor, reactivity loss, the degree of uranium enrichment required for compensating reactivity, and the rate of MA annihilation were calculated. Based on these data, the MA recycling system was set up, and the rate of MA annihilation was evaluated. This is reported. (K.I.)

Simplified compact containment BWR plant

International Nuclear Information System (INIS)

Heki, H.; Nakamaru, M.; Tsutagawa, M.; Hiraiwa, K.; Arai, K.; Hida, T.

2004-01-01

The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to both energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's small power output of 300 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The internal upper entry CRDs reduce the height of the reactor vessel (RPV) and consequently reduce the height of the primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), passive auto catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. The recombiner decreases hydrogen concentration in the PCV in the case of a severe accident. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. The feasibility of CCR safety system has been confirmed by LOCA

Implementación del sistema para municipios, usando pmi

OpenAIRE

Cruz Ruiz, Victor Hugo; Buenaño Vera, Sulay Pamela; Freire Cobo, Lenin

2014-01-01

El objetivo de informe de graduación, es plantear y describir la implementación de un sistema informático en una entidad municipal usando PMI, siguiendo las pautas definidas en la Guía del PMBOK. Nuestro estudio es aplicar la Metodología Administración de Proyectos PMI a un proyecto ya terminado. Para este proyecto hemos tomado la información de la implementación del Sistema CABILDO ERP en el Gobierno Municipal de San Jacinto de Yaguachi, realizado por la empresa ecuatoriana...

Stress corrosion cracking of L-grade stainless steels in boiling water reactor (BWR) plants

International Nuclear Information System (INIS)

Suzuki, Shunichi; Fukuda, Toshihiko; Yamashita, Hironobu

2004-01-01

L-grade stainless steels as 316NG, SUS316L and SUS304L have been used for the BWR reactor internals and re-circulation pipes as SCC resistant materials. However, SCC of the L-grade material components were reported recently in many Japanese BWR plants. The detail investigation of the components showed the fabrication process such as welding, machining and surface finishing strongly affected SCC occurrence. In this paper, research results of SCC of L-grade stainless steels, metallurgical investigation of core shrouds and re-circulation pipings, and features of SCC morphology were introduced. Besides, the structural integrity of components with SCC, countermeasures for SCC and future R and D planning were introduced. (author)

Operation status display and monitoring system for BWR nuclear power plant

International Nuclear Information System (INIS)

Wakabayashi, Yasuo; Hayakawa, Hiroyasu; Kawamura, Atsuo; Kaneda, Mitsunori.

1982-01-01

Lately, the development of the system has been made for BWR plants, which monitors the operating status not only in normal operation but also in abnormal state and also for plant safety. Recently, the improvement of man-machine interface has been tried through the practical use of technique which displays data collectively on a CRT screen relating them mutually. As one of those results, the practical use of an electronic computer and color CRT display for No. 1 unit in the Fukushima No. 2 Nuclear Power Station (2F-1), Tokyo Electric Power Co., is described. Also, new centralized control panels containing such systems were used for the 1100 MWe BWR nuclear power plants now under construction, No. 3 unit of the Fukushima No. 2 Power Station and No. 1 unit of Kashiwazaki-Kariwa Nuclear Power Station (2F-3 and K-1, respectively). The display and monitoring system in 2F-1 plant is the first one in which a computer and color CRTs were practically employed for a BWR plant in Japan, and already in commercial operation. The advanced operating status monitoring system, to which the result of evaluation of the above system was added, was incorporated in the new centralized control panels presently under production for 2F-3 and K-1 plants. The outline of the system, the functions of an electronic computer, plant operating status monitor, surveillance test guide, the automation of plant operation and auxiliary operation guide are reported for these advanced monitoring system. It was confirmed that these systems are useful means to improve the man-machine communication for plant operation minitoring. (Wakatsuki, Y.)

BWR [boiling water reactor] core criticality versus water level during an ATWS [anticipated transient without scram] event

International Nuclear Information System (INIS)

Sehgal, B.R.; Peng, C.M.; Maly, J.

1988-01-01

The BWR [boiling water reactor] emergency procedures guidelines recommend management of core water level to reduce the power generated during an anticipated transient without scram (ATWS) event. BWR power level variation has traditionally been calculated in the system codes using a 1-D [one-dimensional] 2-group neutron kinetics model to determine criticality. This methodology used also for calculating criticality of the partially covered BWR cores has, however, never been validated against data. In this paper, the power level versus water level issues in an ATWS severe accident are introduced and the accuracy of the traditional methodology is investigated by comparing with measured data. It is found that the 1-D 2-group treatment is not adequate for accurate predictions of criticality and therefore the system power level for the water level variations that may be encountered in a prototypical ATWS severe accident. It is believed that the current predictions for power level may be too high

ECP measurements in the BWR-1 water loop relative to water composition changes

Energy Technology Data Exchange (ETDEWEB)

Kus, P.; Vsolak, R.; Kysela, J., E-mail: ksp@ujv.cz [Nuclear Research Inst. Rez plc, Husinec - Rez (Czech Republic); Hanawa, S.; Nakamura, T.; Uchida, S., E-mail: hanawa.satoshi@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan)

2010-07-01

The goal of this study is to investigate the usage of ECP sensors in nuclear power plants. ECP sensors were tested using the LVR-15 reactor at the Nuclear Research Institute Rez plc (NRI) in the Czech Republic. The experiment took place on the BWR-1 loop, which was designed for investigating the behaviour of structural materials and radioactivity transport under BWR conditions. The BWR-1 loop facilitates irradiation experiments within a wide range of operating parameters (max. pressure of 10 MPa, max. temperature of 573 K and a neutron flux of 1.0* 10{sup 18} n/m{sup 2}s). This study involves the measurement of electrochemical potential (ECP). Corrosion potential is the main parameter for monitoring of water composition changes in nuclear power plants (NPP). The electrochemical potentials of stainless steel were measured under high temperatures in a test loop (BWR-1) under different water composition conditions. Total neutron flux was ∼10{sup -3} to ∼10{sup 12} n/cm{sup 2}s (>0.1 MeV) at a temperature of 560K, neutral pH, and water resistivity of 18.2 MOhm. ECP sensor response related to changes in water composition was monitored. Switching from NWC (normal water conditions) to HWC (hydrogen water conditions) was controlled using oxygen dosage. Water chemistry was monitored approx. 50 meters from the active channel. The active channel temperature was maintained within a range of 543 - 561 K from the start of irradiation for the entire duration of the experiment. A total of 24 reference electrodes composed of platinum (Pt), silver/silver chloride (Ag/AgCl) and a zircon membrane containing silver oxide (Ag{sub 2}O) powder were installed inside the active channel of the LVR-15 test reactor. The active channel (Field tube) was divided into four zones, with each zone containing six sensors. A mathematical radiolysis code model was created in cooperation with the Japan Atomic Energy Agency. (author)

Stability monitoring for BWR based on singular value decomposition method using artificial neural network

International Nuclear Information System (INIS)

Tsuji, Masashi; Shimazu, Yoichiro; Michishita, Hiroshi

2005-01-01

A new method for evaluating the decay ratios in a boiling water reactor (BWR) using the singular value decomposition (SVD) method had been proposed. In this method, a signal component closely related to the BWR stability can be extracted from independent components of the neutron noise signal decomposed by the SVD method. However, real-time stability monitoring by the SVD method requires an efficient procedure for screening such components. For efficient screening, an artificial neural network (ANN) with three layers was adopted. The trained ANN was actually applied to decomposed components of local power range monitor (LPRM) signals that were measured in stability experiments conducted in the Ringhals-1 BWR. In each LPRM signal, multiple candidates were screened from the decomposed components. However, decay ratios could be estimated by introducing appropriate criterions for selecting the most suitable component among the candidates. The estimated decay ratios are almost identical to those evaluated by visual screening in a previous study. The selected components commonly have the largest singular value, the largest decay ratio and the least squared fitting error among the candidates. By virtue of excellent screening performance of the trained ANN, the real-time stability monitoring by the SVD method can be applied in practice. (author)

Minimización de una función normal-merit mediante un algoritmo convergente globalmente

Directory of Open Access Journals (Sweden)

Gómez Suárez, M.

1998-01-01

Full Text Available En este trabajo presentamos dos conceptos relacionados con la solución de sistemas de ecuaciones no lineales y con desigualdades. El primer concepto es el de una función normal merit, que resume las propiedades básicas que tienen distintas funciones merit conocidas. El segundo concepto es el de un operador Newtoniano, cuyos valores generalizan el concepto de Hessiana para la función normal merit. Combinando el resultado del método generalizado de Newton con ciertos métodos de primer orden, obtenemos un algoritmo de convergencia global para minimizar funciones normales merit.

CFD analysis for the hydrogen transport in the primary contention of a BWR using the codes OpenFOAM and Gas-Flow; Analisis CFD para el transporte de hidrogeno en la contencion primaria de un reactor BWR usando los codigos OpenFOAM y GasFlow

Energy Technology Data Exchange (ETDEWEB)

Jimenez P, D. A.

2014-07-01

using a limited number of semi-empirical data, and instead, mathematical relationships are used taking into account the various physical phenomena as well the interactions that occur among them, such as heat transfer between the fluid and the solid walls condensation of water vapor on the walls, the turbulent effects in areas of restricted passage, etc. Taking into account these advantages, this study presents a qualitative and quantitative comparison between the CFD codes OpenFOAM and Gas-Flow related to the transport phenomena of Hydrogen and other gases in the primary containment of a BWR reactor. Gas-Flow is a code of commercial license that is well validated, developed in Germany to analyze the transport of gases in nuclear reactor containments. On the other hand, OpenFOAM is an open source CFD code offering several solvers for different phenomena assessments, in this work, the reacting Foam solver is used because it has a strong similarity to the intended application of Hydrogen transport. In this thesis the results obtained using the reacting Foam solver of OpenFOAM for the calculation of transport of Hydrogen are compared with the results of the Gas-Flow code in order to assess if it is feasible to use the open source code OpenFOAM in the case of Hydrogen transport in primary containment of a BWR reactor. Some differences in the qualitative and quantitative results from both codes were found, the differences (with a maximum error rate of 4%) in the quantitative results were found are small and are considered more than acceptable for this type of analysis, moreover, these differences are mainly attributed to the transport models used, mainly because OpenFOAM uses a homogeneous mixture model and Gas-Flow a heterogeneous one. Implementing appropriate solvers in codes like OpenFOAM has the goal to develop own tools that are applicable to the transport of Hydrogen in the primary containment of a BWR reactor and thus, to gain some independence while not relying on

Refuerzo de puentes por cambio de esquema estructural: optimización mediante algoritmo genético

OpenAIRE

Valenzuela, Matías A.; Casas Rius, Joan Ramon

2011-01-01

La presente comunicación entrega un estudio detallado del proceso constructivo y de tesado para el refuerzo de puentes con tipología longitudinal de vigas continuas convirtiéndolos en puentes en arco atirantado (tipo network). Se presentan las etapas básicas propuestas en aspectos de construcción, se establecen las hipótesis del estudio de tesado e implementa el método de tesado a partir de la optimización automatizada mediante algoritmos genéticos, definiendo criterios como: variables de ...

Assessment of two BWR accident management strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Petek, M.

1991-01-01

A recently completed Oak Ridge effort proposes two management strategies for mitigation of the events that might occur in-vessel after the onset of significant core damage in a BWR severe accident. While the probability of such an accident is low, there may be effective yet inexpensive mitigation measures that could be implemented employing the existing plant equipment and requiring only additions to the plant emergency procedures. In this spirit, accident management strategies have been proposed for use of a borated solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and for containment flooding to maintain the core debris within the reactor vessel if injection systems cannot be restored. The proposed strategy for poisoning of the water used for vessel reflood should injection systems be restored after control blade damage has occurred has great promise, using only the existing plant equipment but employing a different chemical form for the boron poison. The dominant BWR severe accident sequence is Station Blackout and without means for mechanical stirring or heating of the storage tank, the question of being able to form the poisoned solution under accident conditions becomes of supreme importance. On the other hand, the proposed strategy for drywell flooding to cool the reactor vessel bottom head and prevent the core and structure debris from escaping to the drywell holds less promise. This strategy does, however, have potential for future plant designs in which passive methods might be employed to completely submerge the reactor vessel under severe accident conditions without the need for containment venting

CFD predictions of standby liquid control system mixing in lower plenum of a BWR

International Nuclear Information System (INIS)

Boyd, Christopher; Skarda, Raymond

2014-01-01

Highlights: • Computational fluid dynamics analysis of BWR lower plenum. • Mixing and stratification of the standby liquid control system injection. • Scoping study highlights the expected flow paths and limitations of experiments. - Abstract: During an anticipated transient without scram (ATWS) scenario in certain boiling water reactor (BWR) systems, a standby liquid control system (SLCS) is used to inject a sodium pentaborate solution into the reactor system in order to quickly shut down (scram) the reactor without the use of the control rods. Some BWR designs utilize a SLCS that injects through a set of nozzles on a vertical pipe in the peripheral region of the lower plenum of the reactor vessel. During the scenario, system water levels are reduced and natural circulation flow rates down through the jet pump nozzles and up into the core are a small fraction of the rated system flow. It is during this period that the SLCS flows are considered. This work outlines some initial scoping studies completed by the staff at the Nuclear Regulatory Commission (NRC). An attempt at benchmarking the computational fluid dynamics (CFD) approach using a set of available test data from a small facility is outlined. Due to our lack of information related to specific details of the facility geometry along with the limited data available from the test, the benchmark exercise produced only a qualitative basis for selecting turbulence models and mesh density. A CFD model simulating a full-scale reactor system is developed for the lower plenum of a representative BWR/4 design and SLCS flows and mixing are studied under a range of flow conditions. The full-scale BWR simulation builds upon the lessons learned from the benchmark exercise. One challenge for this work is the large size of the domain and the relatively small size of the geometric details such as flow passages and gaps. The geometry is simplified to make meshing feasible by eliminating some of the small features. The

Assessment of the fracture toughness of irradiated stainless steel for BWR core shrouds

International Nuclear Information System (INIS)

Carter, R.G.; Gamble, R.M.

2002-01-01

Data from previously performed experiments were collected and evaluated to determine the relationship between fracture toughness and neutron fluence for conditions representative of BWR core shrouds. This relationship together with EPFM (elastic-plastic fracture mechanics) analysis methods similar to those in Appendix K of Section XI of the ASME Code were used to compute margin against failure as a function of neutron fluence for postulated cracks in BWR core shrouds. The results indicate that EPFM analyses can be used for flaw evaluation of core shrouds at fluence levels less than 3.10 21 n/cm 2 (E > 1 MeV). At fluence levels equal to or greater than 3.10 21 n/cm 2 , LEFM (linear-elastic fracture mechanics) analyses should be used with K Ic = 55 MPa-(m) 0.5 . (authors)

Investigations on the thermal-hydraulics of a natural circulation cooled BWR fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Kok, H.V.; Hagen, T.H.J.J. van der; Mudde, R.F. [Delft Univ. of Technology (Netherlands)

1995-09-01

A scaled natural circulation loop facility has been built after the Dodewaard Boiling Water Reactor, which is the only operating natural circulation cooled BWR in the world. The loop comprises one fuel assembly, a riser with a downcomer and a condenser with a cooling system. Freon-12 is used as a scaling liquid. This paper reports on the first measurements done with this facility. Quantities like the circulation flow, carry-under and the void-fraction have been measured as a function of power, pressure, liquid level, riser length, condensate temperature and friction factors. The behavior of the circulation flow can be understood by considering the driving force. Special attention has been paid to the carry-under, which has been shown to have a very important impact on the dynamics of a natural circulation cooled BWR.

International comparison calculations for a BWR lattice with adjacent gadolinium pins

International Nuclear Information System (INIS)

Maeder, C.; Wydler, P.

1984-09-01

The results of burnup calculations for a simplified BWR fuel element with two adjacent gadolinium rods are presented and discussed. Ten complete solutions were contributed by Denmark, France, Italy (3), Japan (3), Switzerland and the UK. Partial results obtained from Poland and the USA are included in an Appendix. (Auth.)

Trend of field data on pipe wall thinning for BWR power plants

International Nuclear Information System (INIS)

Hakii, Junichi; Hiranuma, Naoki; Hidaka, Akitaka

2009-01-01

Strongly motivated by every stakeholder not to repeat Mihama Nuclear Power Station pipe rupture accident in August 2004, JSME Main Committee on Codes and Standards on Power Generation Facilities immediately launched a special task force to develop Rules on Pipe Wall Thinning Management for BWR, PWR and fossil Power Plants respectively. The authors describes the process of the development of Rules for BWR Power Plans from the view point of collections and analysis of fields data of pipe wall thinning. Through its activities, the authors confirmed the existing findings, like the effect of Oxygen injection, turbulence and dependence on coolant temperature, derived from series of laboratory-scaled experiments in FAC and coolant velocities effects in LDI. Further based upon the said proven findings with field data, they explain the adequacy of major concept of the rule such as separate treatment of FAC (Flow Accelerated Corrosion) and LDI (Liquid Droplet Impingement). (author)

BWR - Spent Fuel Transport and Storage with the TNTM9/4 and TNTM24BH Casks

International Nuclear Information System (INIS)

Wattez, L.; Marguerat, Y.; Hoesli, C.

2006-01-01

The Swiss Nuclear Utilities have started in 2001 to store spent fuel in dry metallic dual-purpose casks at ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd., the Muehleberg Nuclear Power Plant owner, is involved in this process and has elected to store its BWR spent fuel in a new high capacity dual-purpose cask, the TNeTeM24BH from the COGEMA Logistics/TRANSNUCLEAR TN TM 24 family. The Muehleberg BWR spent fuels are transported by road in a medium size shuttle transport cask and then transferred to a heavy transport/storage cask (dry transfer) in the hot cell of ZWILAG site. For that purpose, COGEMA Logistics designed and supplied: - Two shuttle casks, TN TM 9/4, mainly devoted to transport of spent fuel from Muehleberg NPP to ZWILAG. Licensed according to IAEA 1996, the TN TM 9/4 is a 40 ton transport cask, for 7 BWR high bum-up spent fuel assemblies. - A series of new high capacity dual-purpose casks, TN TM 24BH, holding 69 BWR spent fuels. Two transport campaigns took place in 2003 and 2004. For each campaign, ten TN TM 9/4 round trips are performed, and one TN TM 24BH is loaded. 5 additional TN TM 24BH are being manufactured for BKW, and the next transport campaigns are scheduled from 2006. The TN TM 24BH high capacity dual purpose cask and the TN TM 9/4 transport cask characteristics and capabilities will then be detailed. (authors)

Applicability of the diffusion and simplified P3 theories for BWR pin-by-pin core analysis

International Nuclear Information System (INIS)

Tada, Kenichi; Yamamoto, Akio; Kitamura, Yasunori; Yamane, Yoshihiro; Watanabe, Masato; Noda, Hiroshi

2007-01-01

The pin-by-pin fine mesh core calculation method is considered as a candidate of next-generation core calculation method for BWR. In this study, the diffusion and the simplified P 3 (SP 3 ) theories are applied to the pin-by-pin core analysis of BWR. Performances of the diffusion and the SP 3 theories for cell-homogeneous pin-by-pin fine mesh BWR core analysis are evaluated through comparison with cell-heterogeneous detailed transport calculation by the method of characteristics (MOC). In this study, two-dimensional, 2x2 multi-assemblies geometry is used to compare the prediction accuracies of the diffusion and the SP 3 theories. The 2x2 multi- assemblies geometry consists of two types of 9x9 UO 2 assembly that have two different enrichment splittings. To mitigate the cell-homogenization error, the SPH method is applied for the pin-by-pin fine mesh calculation. The SPH method is a technique that reproduces a result of heterogeneous calculation by that of homogeneous calculation. The calculation results indicated that diffusion theory shows larger discrepancy than that of SP 3 theory on pin-wise fission rates. Furthermore, the accuracy of the diffusion theory would not be sufficient for the pin-by-pin fine mesh calculation. In contrast to the diffusion theory, the SP 3 theory shows much better accuracy on pin wise fission rates. Therefore, if the SP 3 theory is applied, the accuracy of the pin-by-pin fine mesh BWR core analysis will be higher and will be sufficient for production calculation. (author)

A simplified spatial model for BWR stability

International Nuclear Information System (INIS)

Berman, Y.; Lederer, Y.; Meron, E.

2012-01-01

A spatial reduced order model for the study of BWR stability, based on the phenomenological model of March-Leuba et al., is presented. As one dimensional spatial dependence of the neutron flux, fuel temperature and void fraction is introduced, it is possible to describe both global and regional oscillations of the reactor power. Both linear stability analysis and numerical analysis were applied in order to describe the parameters which govern the model stability. The results were found qualitatively similar to past results. Doppler reactivity feedback was found essential for the explanation of the different regions of the flow-power stability map. (authors)

Level 2 PRA for a German BWR

International Nuclear Information System (INIS)

Sassen, F.; Rapp, W.; Tietsch, W.; Roess, P.

2007-01-01

A concept for a Level 2 Probabilistic Risk Assessment (L2 PRA) for a German Boiling Water Reactor (BWR) has been developed taking into account the role of L2 PRA within the German regulatory landscape. According to this concept, a plant specific evaluation of the severe accident phenomenology as well as analyses of the accident progression for the severe accident scenarios has been performed. Furthermore a plant specific MELCOR 1.8.6 model has been developed and special MELCOR source term calculations have been performed for the different release paths. This paper will present examples from the different areas described above. (author)

Protecting AREVA ATRIUM™ BWR fuel from debris fretting failure

International Nuclear Information System (INIS)

Cole, Steven E.; Garner, Norman L.; Lippert, Hans-Joachim; Graebert, Rüdiger; Mollard, Pierre; Hahn, Gregory C.

2014-01-01

Historically, debris fretting has been the leading cause of fuel rod failure in BWR fuel assemblies, costing the industry millions of dollars in lost generation and negatively impacting the working area of plant site personnel. In this paper the focus will be on recent BWR fuel product innovation designed to eliminate debris related failures. Experience feedback from more than three decades of operation history with non-line-of-sight FUELGUARD™ lower tie plate debris filters will be presented. The development and relative effectiveness of successive generations of filtration technology will be discussed. It will be shown that modern, state of the art debris filters are an effective defense against debris fretting failure. Protective measures extend beyond inlet nozzle debris filters. The comprehensive debris resistance features built into AREVA’s newest fuel design, the ATRIUM™ 11, reduce the overall risk of debris entrapment as well as providing a degree of protection from debris that may fall down on the fuel assembly from above, e.g., during refueling operations. The positive recent experience in a debris sensitive plant will be discussed showing that the combination of advanced fuel technology and a robust foreign material exclusion program at the reactor site can eliminate the debris fretting failure mechanism. (author)

Managing the aging of BWR control rod drive systems

International Nuclear Information System (INIS)

Greene, R.H.; Farmer, W.S.

1992-01-01

This Phase I Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with BWR control and rod drive mechanisms (CRDMs) and assesses the merits of various methods of ''imaging'' this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of the Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the control rod drive (CRD) system, and (4) personal information exchange with nuclear industry CRDM maintenance experts. The report documenting the findings of this research, NUREG-5699, will be published this year. Nearly 23% of the NPRDS CRD system component failure reports were attributed to the CRDM. The CRDM components most often requiring replacement due to aging are the Graphitar seals. The predominant causes of aging for these seals are mechanical wear and thermal embrittlement. More than 59% of the NPRDS CRD system failure reports were attributed to components that comprise the hydraulic control unit (HCU). The predominant HCU components experiencing the effects of service wear and aging are value seals, discs, seats, stems, packing, and diaphragms

Physical model of nonlinear noise with application to BWR stability

International Nuclear Information System (INIS)

March-Leuba, J.; Perez, R.B.

1983-01-01

Within the framework of the present model it is shown that the BWR reactor cannot be unstable in the linear sense, but rather it executes limited power oscillations of a magnitude that depends on the operating conditions. The onset of these oscillations can be diagnosed by the decrease in stochasticity in the power traces and by the appearance of harmonics in the PSD

BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification

International Nuclear Information System (INIS)

Hartmann, C.; Sanchez, V.; Tietsch, W.; Stieglitz, R.

2012-01-01

The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)

VIM Monte Carlo versus CASMO comparisons for BWR advanced fuel designs

International Nuclear Information System (INIS)

Pallotta, A.S.; Blomquist, R.N.

1994-01-01

Eigenvalues and two-dimensional fission rate distributions computed with the CASMO-3G lattice physics code and the VIM Monte Carlo Code are compared. The cases assessed are two advanced commercial BWR pin bundle designs. Generally, the two codes show good agreement in K inf , fission rate distributions, and control rod worths

Development of a detailed BWR core thermal-hydraulic analysis method based on the Japanese post-BT standard using a best-estimate code

International Nuclear Information System (INIS)

Ono, H.; Mototani, A.; Kawamura, S.; Abe, N.; Takeuchi, Y.

2004-01-01

The post-BT standard is a new fuel integrity standard or the Atomic Energy Society of Japan that allows temporary boiling transition condition in the evaluation for BWR anticipated operational occurrences. For application of the post-BT standard to BWR anticipated operational occurrences evaluation, it is important to identify which fuel assemblies and which axial, radial positions of fuel rods have temporarily experienced the post-BT condition and to evaluates how high the fuel cladding temperature rise was and how long the dryout duration continued. Therefore, whole bundle simulation, in which each fuel assembly is simulated independently by one thermal-hydraulic component, is considered to be an effective analytical method. In the present study, a best-estimate thermal-hydraulic code, TRACG02, has been modified to extend it predictive capability by implementing the post-BT evaluation model such as the post-BT heat transfer correlation and rewetting correlation and enlarging the number of components used for BWR plant simulation. Based on new evaluation methods, BWR core thermal-hydraulic behavior has been analyzed for typical anticipated operational occurrence conditions. The location where boiling transition occurs and the severity of fuel assembly in the case of boiling transition conditions such as fuel cladding temperature, which are important factors in determining whether the reuse of the fuel assembly can be permitted, were well predicted by the proposed evaluation method. In summary, a new evaluation method for a detailed BWR core thermal-hydraulic analysis based on the post-BT standard of the Atomic Energy Society of Japan has been developed and applied to the evaluation of the post-BT standard during the actual BWR plant anticipated operational occurrences. (author)

Pre-study of dynamic loads on the internals caused by a large pipe break in a BWR; Foerstudie av stroemningsinducerade laster paa interndelar vid brott i huvudcirkulationskretsarna i BWR

Energy Technology Data Exchange (ETDEWEB)

Marcinkiewicz, Jerzy; Lindgren, Anders [Det Norske Veritas Nuclear Technology AB, Stockholm (Sweden)

2002-12-01

Det Norske Veritas Nuclear Technology has performed a literature study of dynamic load on a BWR (Boiling Water Reactor) internals caused by a large pipe break. The goal of the study was to improve the knowledge about the physics of phenomena occurring in the RPV (Reactor Pressure Vessel) after pipe break in the main circulation system and also to make a review of calculation methods, models and computer programs including their capabilities when calculating the dynamic loads. The report presents description of relevant parts of a BWR, initial and boundary conditions, and phenomena determining the loads - rapid depressurization and propagation of pressure wave (including none-equilibrium). Furthermore, the report generally describes possible methodologies for calculating the dynamic loads on internals after the pipe break and the experiences from calculations the dynamic loads with different methods (computer programs) including comparisons with experimental data. Fluid-Structure Interaction methodology and its importance for calculation of dynamic loads on reactor internals is discussed based on experimental data. A very intensive research program for studying and calculating the dynamic loads on internals after pipe breaks has been performed in USA and Germany during the seventies and the eighties. Several computer programs have been developed and a number of large-scale experiments have been performed to calibrate the calculation methods. In spite of the fact that all experiments were performed for PWR several experiences should be valid also for BWR. These experiences, connected mainly to capabilities of computer programs calculating dynamic loads, are discussed in the report.

Algoritmos GRASP para el equilibrado de líneas con riesgo ergonómico mínimo

OpenAIRE

Bautista Valhondo, Joaquín; Alfaro Pozo, Rocío; Batalla García, Cristina; Llovera Laborda, Sara

2015-01-01

Resumen - Se presenta un problema de equilibrado de líneas de montaje con ciclo compatible y número fijo de estaciones con el objetivo de minimizar el máximo riesgo ergonómico presente en las estaciones de trabajo. Tras la formu lación de un modelo matemático para el problema, se proponen procedimientos GRASP para su resolución. Los procedimientos incorporan un algoritmo greedy para regularizar el riesgo ergonómico así como diferentes tipos de mejora local. De...

An analysis of instabilities of nuclear-coupled density-wave in BWR using modern frequency-domain control theory

International Nuclear Information System (INIS)

Zhao Yangping; Gao Huahun; Fu Longzhou

1991-01-01

A state-of-the-art multi-variable frequency-domain model has been developed for analysis of instabilities of nuclear-coupled density-wave in BWR core. The characteristic locus method is used for analysing the stability of BWR. A computer code-NUCTHIA has been derived. The model has been tested against the existing experimental data and compared with results of past single-variable analyses. By using the NUCTHIA code, the investigations of effects of main system parameters on BWW core stability have also been made. All the results are consistent with the experimental data

BWR containments license renewal industry report; revision 1. Final report

International Nuclear Information System (INIS)

Smith, S.; Gregor, F.

1994-07-01

The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures, and components, in the license renewal technical Industry Reports (IR's). License renewal applicants may choose to reference these IR's in support of their plant-specific license renewal applications as an equivalent to the integrated plant assessment provisions of the license renewal rule (IOCFR54). The scope of the IR provides the technical basis for license renewal for U.S. Boiling Water Reactor (BWR) containments. The scope of the report includes containments constructed of reinforced or prestressed concrete with steel liners and freestanding stell containments. Those domestic BWR containments designated as Mark I, Mark II or Mark III are covered, but no containments are addressed before these designs. The report includes those items within the jurisdictional boundaries for metal and concrete containments defined by Section III of the ASME Boiler and Pressure Vessel Code, Division 1, Subsection NE (Class MC) and Division 2 (Class CC) and their supports, but excluding snubbers

A direct comparison of MELCOR 1.8.3 and MAAP4 results for several PWR ampersand BWR accident sequences

International Nuclear Information System (INIS)

Leonard, M.T.; Ashbaugh, S.G.; Cole, R.K.; Bergeron, K.D.; Nagashima, K.

1996-01-01

This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The PWR system examined in this study is a 1100 MWe system similar in design to a Westinghouse 3-loop plant with a large dry containment; the BWR is a 1100 MWe system similar in design to General Electric BWR/4 with a Mark I containment. A total of nine accident sequences were studied with both codes. Results of these calculations are compared to identify major differences in the timing of key events in the calculated accident progression or other important aspects of severe accident behavior, and to identify specific sources of the observed differences

Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors; Analisis comparativo de resultados entre CASMO, MCNP y SERPENT para una suite de problemas Benchmark en reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Reyes F, M. del C.; Del Valle G, E., E-mail: vicente.xolocostli@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, UP - Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)

2014-10-15

In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

Analysis CFD for the hydrogen transport in the primary containment of a BWR

International Nuclear Information System (INIS)

Jimenez P, D. A.; Del Valle G, E.; Gomez T, A. M.

2014-10-01

This study presents a qualitative and quantitative comparison among the CFD GASFLOW and OpenFOAM codes which are related with the phenomenon of hydrogen transport and other gases in the primary containment of a Boiling Water Reactor (BWR). GASFLOW is a commercial license code that is well validated and that was developed in Germany for the analysis of the gases transport in containments of nuclear reactors. On the other hand, OpenFOAM is an open source code that offers several evaluation solvers for different types of phenomena; in this case, the solver reacting-Foam is used to analyze the hydrogen transport inside the primary containment of the BWR. The results that offer the solver reacting-Foam of OpenFOAM are evaluated in the hydrogen transport calculation and the results are compared with those of the program of commercial license GASFLOW to see if is viable the use of the open source code in the case of the hydrogen transport in the primary containment of a BWR. Of the obtained results so much quantitative as qualitative some differences were identified between both codes, the differences (with a percentage of maximum error of 4%) in the quantitative results are small and they are considered acceptable for this analysis type, also, these differences are attributed mainly to the used transport models, considering that OpenFOAM uses a homogeneous model and GASFLOW uses a heterogeneous model. (Author)

Analysis of radiological consequences in a typical BWR with a mark-II containment

International Nuclear Information System (INIS)

Funayama, Kyoko; Kajimoto, Mitsuhiro

2003-01-01

INS/NUPEC in Japan has been carrying out the Level 3 PSA program. In the program, the MACCS2 code has been extensively applied to analyze radiological consequences for typical BWR and PWR plants in Japan. The present study deals with analysis of effects of the AMs, which were implemented by industries, on radiological consequence for a typical BWR with a Mark-II containment. In the present study, source terms and their frequencies of source terms were used based on results of Level 2 PSA taking into account AM countermeasures. Radiological consequences were presented with dose risks (Sv/ry), which were multiplied doses (Sv) by containment damage frequencies (/ry), and timing of radionuclides release to the environment. The results of the present study indicated that the dose risks became negligible in most cases taking AM countermeasures and evacuations. (author)

Efforts for optimization of BWR core internals replacement

International Nuclear Information System (INIS)

Iizuka, N.

2000-01-01

The core internal components replacement of a BWR was successfully completed at Fukushima-Daiichi Unit 3 (1F3) of the Tokyo Electric Power Company (TEPCO) in 1998. The core shroud and the majority of the internal components made by type 304 stainless steel (SS) were replaced with the ones made of low carbon type 316L SS to improve Intergranular Stress Corrosion Cracking (IGSCC) resistance. Although this core internals replacement project was completed, several factors combined to result in a longer-than-expected period for the outage. It was partly because the removal work of the internal components was delayed. Learning a lesson from whole experience in this project, some methods were adopted for the next replacement project at Fukushima-Daiichi Unit 2 (1F2) to shorten the outage and reduce the total radiation exposure. Those are new removal processes and new welding machine and so on. The core internals replacement work was ended at 1F2 in 1999, and both the period of outage and the total radiation exposure were the same degree as expected previous to starting of this project. This result shows that the methods adopted in this project are basically applicable for the core internals replacement work and the whole works about the BWR core internals replacement were optimized. The outline of the core internals replacement project and applied technologies at 1F3 and 1F2 are discussed in this paper. (author)

Radiation buildup and control in BWR recirculation piping

International Nuclear Information System (INIS)

Meyer, W.; Wood, R.M.; Rao, T.V.; Vook, R.W.

1987-01-01

Boiling water nuclear reactors (BWRs) employ stainless steel (Types 304 or 316 NG) pipes in which high-purity water at temperatures of ∼ 275 0 C are circulated. Various components of the system, such as valves and bearings, often contain hard facing metal alloys such as Stellite-6. These components, along with the stainless steel tubing and feedwater, serve as sources of 59 Co. This cobalt, along with other soluble and insoluble impurities, is carried along with the circulating water to the reactor core where it is converted to radioactive 60 Co. After reentering the circulating water, the 60 Co can be incorporated into a complex corrosion layer in the form of CoCr 2 O 4 and/or CoFe 2 O 4 . The presence of even small amounts of 60 Co on the walls of BWR cooling systems is the dominant contributor to inplant radiation levels. Thus BWR owners and their agents are expending significant time and resources in efforts to reduce both the rate and amount of 60 Co buildup. The object of this research is twofold: (a) to form a thin diffusion barrier against the outward migration of cobalt from a cobalt-containing surface and (b) to prevent the growth of a 60 Co-containing corrosion film. The latter goal was the more important since most of the radioactive cobalt will originate from sources other than the stainless steel piping itself

Knowledge management method for knowledge based BWR Core Operation Management System

Energy Technology Data Exchange (ETDEWEB)

Wada, Yutaka; Fukuzaki, Takaharu; Kobayashi, Yasuhiro

1989-03-01

A knowledge management method is proposed to support an except whose knowledge is stored in a knowledge base in the BWR Core Operation Management System. When the alterations in the operation plans are motivated by the expert after evaluating them, the method attempts to find the knowledge which must be modified and to give the expert guidances. In this way the resultant operation plans are improved by modifying values of referenced data. Using data dependency among data, which are defined and referred during inference, data to be modified are retrieved. In generating modification guidances, data reference and definition procedures are classified by syntactic analysis of knowledge. The modified data values are calculated with a sensitivity between the increment in the data to be modified and the resultant one in the performance of operation plans. The efficiency of the knowledge management by the proposed method, when applied to the knowledge based system including 500 pieces of knowledge for BWR control rod programming, is higher than that for interactive use of existing general purpose editors. (author).

Strain-induced corrosion cracking in ferritic components of BWR primary circuits

International Nuclear Information System (INIS)

Seifert, H.-P.; Ritter, S.; Ineichen, U.; Tschanz, U.; Gerodetti, B.

2003-04-01

The present final report of the RIKORR project is a summary of a literature survey and of the experimental work performed by PSI on the environmentally-assisted cracking (EAC) and dynamic strain ageing (DSA) susceptibility of low-alloy steels (LAS) in high-temperature (HT) water. Within this project, the EAC crack growth behaviour of different low-alloy RPV steels, weld filler and weld heat-affected zone materials has been investigated under simulated transient and steady-state BWR/NWC power operation conditions. The strain-induced corrosion cracking (SICC) / low-frequency corrosion fatigue (CF) and stress corrosion cracking (SCC) crack growth behaviour of different low-alloy RPV steels under simulated transient and stationary BWR/NWC conditions was characterized by slow rising load / low-frequency corrosion fatigue and constant load / periodical partial unloading / ripple load tests with pre-cracked fracture mechanics specimens in oxygenated HT water at temperatures of either 288, 250, 200 or 150 o C. Modern high-temperature water loops, on-line crack growth monitoring and fractographic analysis by scanning electron microscopy (SEM) were used to quantify the cracking response. (author)

Knowledge management method for knowledge based BWR Core Operation Management System

International Nuclear Information System (INIS)

Wada, Yutaka; Fukuzaki, Takaharu; Kobayashi, Yasuhiro

1989-01-01

A knowledge management method is proposed to support an except whose knowledge is stored in a knowledge base in the BWR Core Operation Management System. When the alterations in the operation plans are motivated by the expert after evaluating them, the method attempts to find the knowledge which must be modified and to give the expert guidances. In this way the resultant operation plans are improved by modifying values of referenced data. Using data dependency among data, which are defined and referred during inference, data to be modified are retrieved. In generating modification guidances, data reference and definition procedures are classified by syntactic analysis of knowledge. The modified data values are calculated with a sensitivity between the increment in the data to be modified and the resultant one in the performance of operation plans. The efficiency of the knowledge management by the proposed method, when applied to the knowledge based system including 500 pieces of knowledge for BWR control rod programming, is higher than that for interactive use of existing general purpose editors. (author)

Simulation of hydrogen deflagration and detonation in a BWR reactor building

International Nuclear Information System (INIS)

Manninen, M.; Silde, A.; Lindholm, I.; Huhtanen, R.; Sjoevall, H.

2002-01-01

A systematic study was carried out to investigate the hydrogen behaviour in a BWR reactor building during a severe accident. BWR core contains a large amount of Zircaloy and the containment is relatively small. Because containment leakage cannot be totally excluded, hydrogen can build up in the reactor building, where the atmosphere is normal air. The objective of the work was to investigate, whether hydrogen can form flammable and detonable mixtures in the reactor building, evaluate the possibility of onset of detonation and assess the pressure loads under detonation conditions. The safety concern is, whether the hydrogen in the reactor building can detonate and whether the external detonation can jeopardize the containment integrity. The analysis indicated that the possibility of flame acceleration and deflagration-to-detonation transition (DDT) in the reactor building could not be ruled out in case of a 20 mm 2 leakage from the containment. The detonation analyses indicated that maximum pressure spike of about 7 MPa was observed in the reactor building room selected for the analysis

Control Adaptativo Fraccionario Optimizado por Algoritmos Genéticos, Aplicado a Reguladores Automáticos de Voltaje

Directory of Open Access Journals (Sweden)

Marco E. Ortiz-Quisbert

2016-10-01

Full Text Available Resumen: En este trabajo se presenta la técnica del control adaptable de orden fraccionario por modelo de referencia (CAOFMR, aplicada a los reguladores automáticos de voltaje (RAV. El artículo se enfoca en el ajuste de las ganancias adaptables y los órdenes de derivación de las leyes de ajuste del controlador CAOFMR, determinados por la minimización de una función criterio definida para el modelo simplificado del RAV, mediante la utilización de la técnica de optimización de algoritmos genéticos (AG. En base a un criterio de evaluación propuesto por otros autores, se realizan comparaciones, por medio de simulaciones, de la técnica de control propuesta con los resultados obtenidos por la técnica de control PID de orden entero (OEPID (Zamani et al., 2009. Se muestra que el controlador CAOFMR con parámetros optimizados por AG, entrega mejores resultados en términos de robustez frente a variaciones en los parámetros del sistema controlado y mejoras en relación a la velocidad de convergencia hacia las señales de referencia del sistema RAV. Abstract: The technique Fractional Order Model Reference Adaptive Control (FOMRAC applied to an Automatic Voltage Regulator (AVR is presented in this paper. The work is focused on tuning the adaptive gains and the derivation order of the adaptive laws of the FOMRAC, determined through the minimization of a criterion function defined for the simplified model of the AVR, by means of the genetic algorithm (GA optimization technique. Based on the criterion function proposed by other authors a simulated comparative study is performed, comparing the proposed methodology with the integer order PID control reported in (Zamani et al., 2009. It is shown that the FOMRAC with parameters optimized by GA provides better results in terms of robustness under parameters variations of the system under control and improvements in the convergence speed of the control error. Palabras clave: Control Adaptativo de

ASIGNACIÓN DE SUPERVISORES FORESTALES: RESOLUCIÓN MEDIANTE UN ALGORITMO TABU SEARCH ASSIGNMENT OF FOREST SUPERVISORS: RESOLUTION BY MEANS OF A TABU SEARCH ALGORITHM

Directory of Open Access Journals (Sweden)

Lorena Pradenas Rojas

2008-12-01

Full Text Available En este estudio se presenta un modelo matemático para un problema genérico de asignación de personal. Se implementa y evalúa un procedimiento de solución mediante la metaheurística Tabu Search. El algoritmo propuesto es usado para resolver un caso real de asignación de supervisores forestales. Los resultados muestran que el algoritmo desarrollado es eficiente en la resolución de este tipo de problema y tiene un amplio rango de aplicación para otras situaciones reales.This study presents a mathematical model for a generic problem of staff allocation. A solution is implemented and evaluated by means of the Tabu Search metaheuristic. The proposed algorithm is used to solve a real case of forestry supervisors' allocation. The results show that the developed algorithm is efficient solving this kind of problems and that it has a wide range of application for other real situations.

Evaluación del comportamiento de un algoritmo para la secuenciación en un taller de flujo con tiempos dependientes de la secuencia

Directory of Open Access Journals (Sweden)

Adriano da Silva-Carvalho

2010-01-01

Full Text Available En el trabajo se realiza la evaluación del comportamiento de un algoritmo propuesto para la secuenciación en un taller de flujo con tiempos dependientes de la secuencia, utilizando un método híbrido basado en el Algoritmo de Búsqueda del Extremo de una Función de un Código Variable y la metaheurística de Búsqueda Dispersa. Asimismo, se analizan casos obtenidos de la aplicación en un taller mecánico de una industria nacional.In this paper the evaluation of the behaviour of a proposed algorithm is made for the sequence of events in a stream flow in a workshop with dependent moments of the sequence using a hybrid method based on the SEARCH OF THE EXTREME OF A FUNTION OF A VARIABLE CODE ALGORITHM AND THE DISPERSED SEARCH METHOD ALGORITHM In the same way, cases obtained from the application in a workshop of the national mechanical industry are analysed.