WorldWideScience

Sample records for reactor post-procesador para

  1. El método de los elementos finitos para el modelado de ondas con un procesador vectorial

    OpenAIRE

    Sanz, F.; Serón, Francisco J.; Kindelan, M.; Pérez, C.

    1990-01-01

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

  2. 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...

  3. Modelación y simulación de disipadores de calor para procesadores de computadora en COMSOL Multiphysics Modeling and simulation of heat sinks for computer processors in COMSOL Multiphysics

    Directory of Open Access Journals (Sweden)

    Sulin Garro Acón

    2012-11-01

    Full Text Available En este estudio se analizó la transferencia de calor en tres disipadores de calor utilizados para enfriar los procesadores de computadoras de escritorio. El objetivo de estos disipadores es evitar el sobrecalentamiento de la unidad de procesamiento y la consecuente reducción de la vida útil del computador. Los disipadores de calor se modelaron usando COMSOL Multiphysics con las dimensiones reales de los dispositivos y la generación de calor se modeló con una fuente puntual. Luego se modificaron los diseños de los disipadores para lograr una temperatura más baja en la zona más caliente del procesador. El resultado fue una reducción en la temperatura en el rango de 5-78 grados Kelvin, al rediseñarse el disipador de calor con variaciones feasibles como la reducción del grosor de las placas de intercambio de calor y el aumento de su número. Esto demuestra la posibilidad de desarrollar diseños optimizados para disipadores de calor que no requieran más materiales sino una mejor ingeniería. El trabajo se inició como parte del curso CM-4101 Modelización y Simulación.In this study, the heat transfer of three desktop- computer heat sinks was analyzed. The objective of using these heat sinks is to avoid overheating of the computer’s processing unit and in turn reduce the corresponding loss in the unit’s service time. The heat sinks were modeled using COMSOL Multiphysics with the actual dimensions of the devices, and heat generation was modeled with a point source. In the next step, the heat sink designs were modified to achieve a lower temperature in the higher temperature location on the heat sink. The results were temperature reductions in the range of 5-78 degrees Kelvin, by making feasible variations in design such as reducing the thickness of the heat exchanger fins and increasing their number. This paper demonstrates that there is room to develop improved designs that do not require more materials but rather a better engineering

  4. Modelizacion, control e implementacion de un procesador energetico paralelo para aplicacion en sistemas multisalida

    Science.gov (United States)

    Ferreres Sabater, Agustin

    Cualquier sistema electronico que incluya un procesado o tratamiento de la senal, y ademas, algun tipo de actuador mecanico generalmente necesita, como minimo, dos tensiones diferentes de alimentacion. Excluyendo los sistemas de alimentacion distribuida, la solucion tecnica mas utilizada para proporcionar dos o mas tensiones consiste en las fuentes de alimentacion multisalida. En una fuente de alimentacion multisalida los diferentes circuitos que conforman cada salida comparten un mismo transformador de potencia optimizando coste, masa, y volumen. Las ventajas obtenidas con este procedimiento tienen en su contra el efecto que sobre cada salida individual provocan las demas en su conjunto debido, principalmente, a los efectos de los elementos parasitos de los componentes. Un cambio de carga en una de las salidas produce un transitorio que es visto por todas las demas como un efecto de impedancia cruzada, y al final del transitorio, la tension de cada salida es diferente respecto a la que tenian antes del transitorio. Este ultimo resultado se conoce como regulacion cruzada. La disminucion de los efectos de la regulacion cruzada ha sido objeto de estudio durante los ultimos anos. El objetivo ha sido el desarrollo de distintas estrategias que permiten, desde disminuir los efectos de la regulacion cruzada hasta los niveles deseables, a eliminarla completamente. El resultado final suele suponer una penalizacion sobre el diseno del sistema directamente proporcional al grado de regulacion a conseguir en las distintas salidas. Entre las soluciones propuestas para eliminar la regulacion cruzada las tecnicas de post-regulacion se han consolidado como la opcion mas aceptada ya que, pueden aplicarse a cualquier convertidor y no suponen ninguna complejidad adicional a la hora de plantear el diseno. En esta Tesis Doctoral se abordara el estudio de la tecnica conocida como postregulacion mediante transformador controlado, que si bien se ha empleado en convertidores resonantes, su

  5. Diseño del núcleo de un procesador AVR de 8 bits utilizando lógica programable de Altera

    Directory of Open Access Journals (Sweden)

    Dilaila Criado

    2010-09-01

    Full Text Available Normal 0 21 false false false ES-TRAD X-NONE X-NONE MicrosoftInternetExplorer4 /* Style Definitions */ table.MsoNormalTable {mso-style-name:"Tabla normal"; mso-tstyle-rowband-size:0; mso-tstyle-colband-size:0; mso-style-noshow:yes; mso-style-priority:99; mso-style-qformat:yes; mso-style-parent:""; mso-padding-alt:0cm 5.4pt 0cm 5.4pt; mso-para-margin:0cm; mso-para-margin-bottom:.0001pt; mso-pagination:widow-orphan; font-size:11.0pt; font-family:"Calibri","sans-serif"; mso-ascii-font-family:Calibri; mso-ascii-theme-font:minor-latin; mso-fareast-font-family:"Times New Roman"; mso-fareast-theme-font:minor-fareast; mso-hansi-font-family:Calibri; mso-hansi-theme-font:minor-latin; mso-bidi-font-family:"Times New Roman"; mso-bidi-theme-font:minor-bidi;} En este trabajo se describe, el diseño del núcleo de un microcontrolador AVR, en específico uno similar al AT90S8515. Se utiliza el lenguaje de descripción de hardware VHDL con dispositivos programables de las familias Cyclone y Cyclone II de Altera. Los objetivos que se persiguen con este trabajo son; valorar alternativas de diseño de sistemas digitales en circuitos programables, contar con el diseño de un núcleo AVR para futuras aplicaciones como modulo IP y realizar diseños de sistemas con varios procesadores (SoPC.

  6. El modelo de inventarios de mercancías considerando la interacción entre procesadores y especuladores

    OpenAIRE

    César Revoredo Giha

    2008-01-01

    Este artículo considera una versión alternativa del modelo de inventarios de mercancías bajo expectativas racionales, donde tanto especuladores como firmas procesadoras almacenan productos. El modelo identifica tanto los inventarios llevados por los procesadores como aquellos descritos por el modelo de oferta de inventarios. Sin embargo, en lugar de usar el motivo de conveniencia (retornos por conveniencia), la demanda por inventarios es derivada a partir de un modelo microeconómico de invent...

  7. Extensión del lenguaje y modelo Simplesem con soporte para paralelismo

    Directory of Open Access Journals (Sweden)

    Diez de Medina, Lucas L.

    2013-12-01

    Full Text Available El modelo de ejecución planteado por Carlo Ghezzi y Mehdi Jazayeri, conocido como Simplesem, no contempla lenguajes con instrucciones de ejecución paralela. El objectivo es extender esta herramienta educativa incorporando al modelo la existencia de múltiples procesadores primitivas de lenguaje que permitan expresar conceptos básicos de paralelismo. Para ello, se desarrolló una herramienta que permite analizar de manera gráfica y sencilla el impacto de programas multihilo sobre instrucciones de bajo nivel, que operan directamente sobre la memoria compartida de una máquina virtual. Esta extensión permite representar la semántica operacional de lenguajes paralelos que requieren procesadores multihilo, comunes en la actualidad. Se pretende distribuir esta herramienta a la comunidad educativa, de manera tal que puedan realizarse estudios sobre los beneficios obtenidos al aplicarla durante el proceso de aprendizaje de los lenguajes de programación.

  8. Uso de detectores de neutrinos para el monitoreo de reactores nucleares Uso de detectores de neutrinos para el monitoreo de reactores nucleares

    Directory of Open Access Journals (Sweden)

    Gerardo Moreno

    2012-02-01

    Full Text Available Se estudia la factibilidad del uso de los detectores de antineutrinos para el monitoreo de reactores nucleares. Usando un modelo sencillo de cascada de fisión a dos componentes, se ilustra la dependencia del número de antineutrinos detectados a una distancia L del reactor según la composición nuclear del combustible. Se explica el principio de detección de neutrinos de reactores en base al decaimiento beta inverso y se describe como los detectores de neutrinos pueden emplearse para el monitoreo de la producción de materiales fisibles en el reactor. Se comenta como generalizar este análisis al caso real de un reactor nuclear in situ y uno de los principales experimentos internacionales dedicados a este propósito. We study the feasibility to use antineutrinos detectors for monitoring of nuclear reactors. Using a simple model of fission shower with two components, we illustrate how the numbers of antineutrinos detected at a distance L from the reactor depend on the composition of the nuclear combustible. We explain the principles of reactor neutrino detection using inverse beta decays and we describe how neutrinos detectors can be used for monitoring the production of fissile materials within the reactors. We comment how to generalize this analysis to the realistic case of a nuclear reactor in situ and one of the main international experiments dedicated to study the use of neutrinos detectors as nuclear safeguards.

  9. Técnicas de gamificación para la motivación de los estudiantes

    OpenAIRE

    Morro González, Sofía

    2015-01-01

    El objetivo de este trabajo es desarrollar el módulo central para el futuro sistema web de la asignatura Procesadores de Lenguajes impartida en la Escuela Técnica Superior de Ingenieros Informáticos de la Universidad Politécnica de Madrid. Para dicho desarrollo se aplicarán técnicas de gamificación con el objetivo de mejorar el aprendizaje, subir la tasa de aprobados de la asignatura y despertar interés por parte del alumnado de la misma. Se modelizaron, diseñaron e implemen...

  10. Método y sistema de modelado de memoria cache

    OpenAIRE

    Posadas Cobo, Héctor; Villar Bonet, Eugenio; Díaz Suárez, Luis

    2010-01-01

    Un método de modelado de una memoria cache de datos de un procesador destino, para simular el comportamiento de dicha memoria cache de datos en la ejecución de un código software en una plataforma que comprenda dicho procesador destino, donde dicha simulación se realiza en una plataforma nativa que tiene un procesador diferente del procesador destino que comprende dicha memoria cache de datos que se va a modelar, donde dicho modelado se realiza mediante la ejecución en dicha plataforma nativa...

  11. Effect of post-digestion temperature on serial CSTR biogas reactor performance

    DEFF Research Database (Denmark)

    Boe, Kanokwan; Karakashev, Dimitar Borisov; Trably, Eric

    2009-01-01

    The effect of post-digestion temperature on a lab-scale serial continuous-flow stirred tank reactor (CSTR) system performance was investigated. The system consisted of a main reactor operated at 55 degrees C with hydraulic retention time (HRT) of 15 days followed by post-digestion reactors with HRT...

  12. Efficient hardware implementation of a full COFDM processor with robust channel equalization and reduced power consumption

    Directory of Open Access Journals (Sweden)

    Alexander López Parrado

    2013-01-01

    Full Text Available Este trabajo presenta el diseño de un procesador banda-base para multiplexación por división de frecuencias ortogonales codificada (COFDM de 12 Mb/s para el estándar IEEE 802.11a. El procesador COFDM banda- base fue diseñado usando circuitos diseñados por nosotros para corrección de fase de portadora, sincronización de tiempo de símbolo, ecualización de canal robusta y decodificación Viterbi. Estos circuitos son flexibles, parametrizados y descritos usando VHDL estructural y genérico. El procesador COFDM banda- base tiene dos dominios de reloj para reducción del consumo de potencia, fue sintetizado sobre un FPGA Stratix II y fue probado experimentalmente usando circuitería de radio frecuencia (RF a 2.4 GHz.

  13. Control para un equipo de análisis químico por RMN basado en un microprocesador de 32 bits

    OpenAIRE

    Palacio Tárrega, Víctor Manuel

    2009-01-01

    El objetivo del PFC es desarrollar el sistema de control de un equipo de análisis quimico por RMN. Las especificacions del sistema de control implican que este integre los ortos dos componentes del equipo de análisis ( procesado de señal y sistema de raiofrecuencia ) con una interfaz gráfica para el usuario.Se desarrollan dos componentes para el sistema de control, hardware y software, El hardware de control incrustado está basado en un procesador ARM de 32 bits,, y se completado con todos lo...

  14. Procesadores de Información : Una Tecnología blanda para el docente

    Directory of Open Access Journals (Sweden)

    Alfonso Orantes

    1993-06-01

    Full Text Available En este trabajo se presenta una aplicación dealgunos recursos para procesar la información que seha desarrollado dentro de la Psicología de la Instrucciónde fácil acceso al docente para facilitar su trabajo y el aprendizaje del estudiante. Estos recursosestán dirigidos a estimular la participación activa delaprendiz. Representan una tecnología blanda, al alcance,medida y comprensión del docente con ventajassobre las tecnologías convencionales apoyadas enaparatos que no se ajustan a nuestra ideosincrasia nirecursos. Se describe como contexto los factores queinfluyen en el aprendizaje de un texto (Aprendiz,Materiales, Actividades y la Dificultad de la Tarea.Asimismo se consideran factores como demandasde tarea, las estrategias de enseñanza y deaprendizaje y se describen sucintamente algunostrabajos venezolanos que han explorado variablesmoderadoras que afectan la eficiencia de este tipo deayudas.

  15. Post-processor for simulations of the ORIGEN program and calculation of the composition of the activity of a burnt fuel core by a BWR type reactor; Post-procesador para simulaciones del programa ORIGEN y calculo de la composicion de la actividad de un nucleo de combustible quemado por un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Sandoval V, S. [IIE, Av. Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico)]. e-mail: sandoval@iie.org.mx

    2006-07-01

    The composition calculation and the activity of nuclear materials subject to processes of burnt, irradiation and decay periods are of utility for diverse activities inside the nuclear industry, as they are it: the processes design and operations that manage radioactive material, the calculation of the inventory and activity of a core of burnt nuclear fuel, for studies of type Probabilistic Safety Analysis (APS), as well as for regulation processes and licensing of nuclear facilities. ORIGEN is a program for computer that calculates the composition and the activity of nuclear materials subject to periods of burnt, irradiation and decay. ORIGEN generates a great quantity of information whose processing and analysis are laborious, and it requires thoroughness to avoid errors. The automation of the extraction, conditioning and classification of that information is of great utility for the analyst. By means of the use of the post-processor presented in this work it is facilitated, it speeds up and wide the capacity of analysis of results, since diverse consultations with several classification options and filtrate of results can be made. As illustration of the utility of the post-processor, and as an analysis of interest for itself, it is also presented in this work the composition of the activity of a burned core in a BWR type reactor according to the following classification criteria: by type of radioisotope (fission products, activation products and actinides), by specie type (gassy, volatile, semi-volatile and not volatile), by element and by chemical group. The results show that the total activity of the studied core is dominated by the fission products and for the actinides, in proportion four to one, and that the gassy and volatile species conform a fifth part of the total activity of the core. (Author)

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

    Directory of Open Access Journals (Sweden)

    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.

  17. Gestión del tiempo ocioso dinámico para ajustar el consumo de energía en tareas de tiempo real integrando control multifrecuencia

    Directory of Open Access Journals (Sweden)

    Alfonso Salvador Alfonsi Sebastiani

    2014-11-01

    Full Text Available Este trabajo tiene como objetivo ajustar la energía consumida por tareas de tiempo real críticas, producto de la variabilidad de sus tiempos de cómputo usando la integración del control multifrecuencia y la planificación realimentada. Se adaptaron e integraron técnicas dinámicas que manejan el tiempo ocioso debido al tiempo de ejecución en el peor caso, al factor de carga del procesador y el aprovechamiento del tiempo ocioso por estiramiento a la(s próxima(s activación(es, en una Técnica Dinámica Multifrecuencia para el Manejo del Tiempo Ocioso, alojada en un planificador realimentado para el ahorro de energía, dirigido a procesadores que varían el voltaje de alimentación y frecuencia de operación. Además se tomó ventaja de las técnicas de control multifrecuencia dado que la gestión de recursos es formulada como un problema de control de sistemas de cómputo que especifica a cada tarea por un lazo de control que trabaja a su propio periodo de activación, diferentes a los requeridos en la referencia y respuesta del sistema (factor de carga total del procesador.Los resultados arrojan que el tiempo ocioso debido a variabilidad de los tiempos de cómputo se distribuye de forma natural por los lazos de control multifrecuencia, global y localmente, pudiendo llegar a un ahorro de energía del 61,04%, dando un valor agregado Intra e InterTarea. Además, sugiere un buen desempeño al contrastarlo con otras estrategias.

  18. 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 ...

  19. Diseño y construcción de un prototipo digital para diagnósticar fallas en motores de inducción

    Directory of Open Access Journals (Sweden)

    FERNANDO VILLADA DUQUE

    2007-01-01

    Full Text Available En este trabajo se presenta el desarrollo de un prototipo digital utilizando un procesador de señales digitales DSP, al cual se le han incorporado dos algoritmos para diagnosticar fallas en el estator de motores de inducción. El primer algoritmo utiliza las redes neuronales artificiales para estimar la corriente de secuencia negativa, la cual es utilizada como indicador de falla. El segundo algoritmo utiliza la impedancia de secuencia inversa como indicador de falla. Se presenta la estructura general del prototipo y su implementación. Se incluyen los resultados obtenidos experimentalmente en un motor de 3 HP utilizando medida en línea a través del prototipo digital

  20. Post-Irradiation Examination and In-Pile Measurement Techniques for Water Reactor Fuels

    International Nuclear Information System (INIS)

    2009-12-01

    Today nuclear technology plays an increasingly important role in our everyday lives, i.e. in energy, industry, medical and environmental applications. Faced with the immediate world's problems of economics, greenhouse gas emissions and water scarcity, as well as the future demand for electricity, nuclear power would provide a long term solution. However better reactor design is required to fulfil such objectives. Therefore, after its stagnation, nuclear engineering has been going through a revival which is reflected in the start of such international projects as Generation IV, INPRO, GNEP and others. These development programs include consideration of a wide range of nuclear reactors of different types and purposes, from high temperature gas cooled and fast reactors with different coolant options to thermal water cooled ones which have both enhanced operating safety and efficient operation due to the optimal design and coolant parameters, etc. Requirements for enhanced reactor safety and efficiency make it necessary to perform precise in- and post-reactor experiments and, consequently, to use more up-to-date measurement equipment and analysis techniques, thus developing hot labs and research reactor facilities. Application of new techniques for measurement and analysis is also related to the consideration of advanced materials for future innovative nuclear reactors with challenging operational conditions that differ greatly from those of the existing nuclear reactors. The necessity to use the most up-to-date precise equipment follows from the necessity to prolong the operating lifetime of the existing NPPs. The designed lifetime of units of many NPPs under operation is practically over. Since these units are in satisfactory condition and the construction of new NPPs is very expensive, it is reasonable to justify their lifetime more precisely and prolong it. However, it requires additional in- and post-reactor examinations. The majority of the hot labs were designed

  1. Research reactor fuel bundle design review by means of hydrodynamic testing; Ensayos hidrodinamicos para verificacion de diseno de un elemento combustible para reactores de investigacion

    Energy Technology Data Exchange (ETDEWEB)

    Pastorini, A; Belinco, C [Comision Nacional de Energia Atomica, San Martin (Argentina). Centro Atomico Constituyentes

    1998-12-31

    During the design steps of a fuel bundle for a nuclear reactor, some vibration tests are usually necessary to verify the prototype dynamical response characteristics and the structural integrity. To perform these tests, the known hydrodynamic loop facilities are used to evaluate the vibrational response of the bundle under the different flow conditions that may appear in the reactor. This paper describes the tests performed on a 19 plate fuel bundle prototype designed for a low power research reactor. The tests were done in order to know the dynamical characteristics of the plates and also of the whole bundle under different flow rate conditions. The paper includes a description of the test facilities and the results obtained during the dynamical characterization tests and some preliminary comments about the tests under flowing water are also presented. (author) 4 refs., 12 figs., 4 tabs. [Espanol] Durante el diseno de un elemento combustible para un reactor nuclear se requiere de la realizacion de ensayos con el objeto de verificar el comportamiento de ese diseno y permitir, de ser necesario, la introduccion de modificaciones al mismo. Para verificar las caracteristicas de respuesta dinamica e integridad estructural, se realizan ensayos de vibraciones que incluyen someter al prototipo a condiciones de circulacion del fluido similares a las que soportara durante la operacion del reactor. Estos ensayos se realizan en facilidades de ensayos conocidas como circuitos hidrodinamicos, que permiten no solo someter el prototipo al flujo de fluido, sino tambien obtener una adecuada caracterizacion de la respuesta del mismo a traves del luso de sensores de distinto tipo. En este trabajo se describen los ensayos realizados sobre un prototipo de elemento combustible de 19 placas destinado a un reactor de investigacion multiproposito de baja potencia. Los ensayos tuvieron como objetivo conocer la respuesta dinamica de las placas individuales y del elemento combustible en su

  2. Active species in a large volume N2-O2 post-discharge reactor

    International Nuclear Information System (INIS)

    Kutasi, K; Pintassilgo, C D; Loureiro, J; Coelho, P J

    2007-01-01

    A large volume post-discharge reactor placed downstream from a flowing N 2 -O 2 microwave discharge is modelled using a three-dimensional hydrodynamic model. The density distributions of the most populated active species present in the reactor-O( 3 P), O 2 (a 1 Δ g ), O 2 (b 1 Σ g + ), NO(X 2 Π), NO(A 2 Σ + ), NO(B 2 Π), NO 2 (X), O 3 , O 2 (X 3 Σ g - ) and N( 4 S)-are calculated and the main source and loss processes for each species are identified for two discharge conditions: (i) p = 2 Torr, f = 2450 MHz, and (ii) p = 8 Torr, f = 915 MHz; in the case of a N 2 -2%O 2 mixture composition and gas flow rate of 2 x 10 3 sccm. The modification of the species relative densities by changing the oxygen percentage in the initial gas mixture composition, in the 0.2%-5% range, are presented. The possible tuning of the species concentrations in the reactor by changing the size of the connecting afterglow tube between the active discharge and the large post-discharge reactor is investigated as well

  3. Sterilization of E. coli bacterium in a flowing N2-O2 post-discharge reactor

    International Nuclear Information System (INIS)

    Villeger, S; Cousty, S; Ricard, A; Sixou, M

    2003-01-01

    Effective destruction of Escherichia coli (E. coli) bacteria has been obtained in a flowing N 2 -O 2 microwave post-discharge reactor. The sterilizing agents are the O atoms and the UV emissions of NOβ which are produced by N and O atoms recombination in the reactor. In the following plasma conditions: pressure 5 Torr, flow rate 1 L n min -1 , microwave power of 100 W in a quartz tube of 5 mm, an O atom density of 2.5x10 15 cm -3 is measured by NO titration in the post-discharge reactor with UV emission in a N 2 -(5%)O 2 gas mixture. Full destruction of 10 13 cfu ml -1 E. coli is observed after a treatment time of 25 min. (rapid communication)

  4. LMRE: Un entorno multiprocesador para la enseñanza de conceptos de concurrencia en un curso CS1

    Directory of Open Access Journals (Sweden)

    De Giusti, Laura Cristina

    2012-01-01

    Full Text Available Se presenta un entorno visual interactivo para la enseñanza de conceptos de concurrencia y paralelismo en un curso inicial de algoritmos. El entorno LMRE (Lidi MultiRobot Environment es una evolución del Visual Da Vinci utilizado extensamente en la introducción a la programación en varias Universidades. El artículo analiza la problemática del cambio tecnológico a partir de la introducción de los procesadores de múltiples núcleos y su impacto sobre la programación y describe una definición del entorno, así como las primitivas a utilizar en la programación de aplicaciones concurrentes. Por último se detallan aspectos de implementación del prototipo actualmente en prueba, así como la evolución del mismo para ser empleado en cursos más avanzados de concurrencia.

  5. Estudio de un reactor catalítico para la obtención de gas de síntesis

    OpenAIRE

    Romero Sayago, Sara Isabel

    2016-01-01

    Este trabajo se centra en el estudio del proceso de reformado de gas natural con vapor de agua para producir gas de síntesis. Un compuesto, que como su nombre indica, es de gran importancia en la síntesis de muchos productos. En concreto, se estudia el reactor heterogéneo catalítico donde tiene lugar la reacción de reformado. Mediante un programa de simulación de procesos químicos, se optimiza el proceso de reformado para obtener un rendimiento elevado en el reactor con el mínimo consumo e...

  6. FPGAs Implementation of fast algorithms oriented to mp3 audio decompression

    Directory of Open Access Journals (Sweden)

    Antonio Benavides

    2012-01-01

    Full Text Available La ejecución de los algoritmos de descompresión de audio exige procesadores potentes con alto nivel de desempeño, sin embargo, dichos algoritmos no son apropiados para aplicaciones óptimas en dispositivos móviles. En este trabajo se lleva a cabo una exploración de algunos algoritmos cuya implementación en hardware permite mejorar el desempeño de los procesadores usados en dispositivos móviles que ejecutan tareas de descompresión de audio. Se presentan algunos resultados experimentales y análisis comparativos.

  7. Classificação de árvores de eucalipto para postes em sistema agroflorestal

    Directory of Open Access Journals (Sweden)

    Daniel de Paula Silveira

    2011-08-01

    Full Text Available Foram construídas tabelas de dupla entrada para quantificação de postes de eucalipto em sistema agroflorestal (SAF, seguindo-se as normas Light, ABPM-E86 e CEMIG-ABNT. Além de permitir definição do melhor tipo de poste com base no seu comprimento máximo, ainda com a árvore em pé, as tabelas construídas podem ser empregadas em inventários de postes de eucalipto. Para a construção das tabelas foram utilizados dados de 114 árvores-amostra de Eucalyptus camaldulensis Dehnh. abatidas e cubadas em um sistema agroflorestal, em que os diâmetros e as alturas variavam de 17 a 43 cm e de 19 a 39 m, respectivamente.

  8. Soc para la identificación de variaciones morfológicas del eritrocito

    Directory of Open Access Journals (Sweden)

    Danelia Matos Molina

    2011-03-01

    Full Text Available Normal 0 21 false false false MicrosoftInternetExplorer4 st1:*{behavior:url(#ieooui } /* Style Definitions */ table.MsoNormalTable {mso-style-name:"Tabla normal"; mso-tstyle-rowband-size:0; mso-tstyle-colband-size:0; mso-style-noshow:yes; mso-style-parent:""; mso-padding-alt:0cm 5.4pt 0cm 5.4pt; mso-para-margin:0cm; mso-para-margin-bottom:.0001pt; mso-pagination:widow-orphan; font-size:10.0pt; font-family:"Times New Roman"; mso-ansi-language:#0400; mso-fareast-language:#0400; mso-bidi-language:#0400;} En el presente se expone el diseño e implementación de un Sistema on Chip apropiado para el procesamiento de señales e imágenes. Este sistema puede representar parte importante en dispositivos médicos de apoyo en el diagnóstico de ciertas  enfermedades donde ocurran variaciones morfológicas en los eritrocitos, como es el caso de la Siklemia. El diseño está basado en un SoC de Plasma y un procesador MIPS Lite, se ha añadido además un núcleo FFT para el mejoramiento de los cálculos de la Transformada discreta de Fourier. Igualmente ha sido desarrollado un programa que calcula dada una imagen, la FFT de los bordes de los eritrocitos presentes.

  9. TIEMPO RECOMENDADO PARA UNA NUEVA CONCEPCIÓN POST ABORTO ESPONTÁNEO

    OpenAIRE

    Baltra E,Estebeni; de Mayo G,Tomás; Rojas G,María de los Ángeles; Arraztoa V,José Antonio

    2008-01-01

    Antecedentes: La recomendación del clínico acerca del tiempo a esperar para una nueva concepción post aborto espontáneo correspondería a una práctica basada en la experiencia y no en la evidencia. Objetivo: Análisis crítico de la literatura científica, en relación al tiempo de espera para intentar un nuevo embarazo en pacientes con aborto espontáneo, y los resultados materno-perinatales asociados a las diferentes conductas. Búsqueda sistemática en múltiples bases de datos. Resultados: Se enco...

  10. Investigation of neutron irradiated reactor vessel steels using post-irradiation annealing techniques

    Energy Technology Data Exchange (ETDEWEB)

    Nakata, Hayato; Fukuya, Koji [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    The matrix damage is known to be a major factor that contributes to embrittlement and hardening of irradiated reactor vessel steels, and is assumed to be composed of the point defect clusters. However field emission gun scanning transmission electron microscopy (FEGSTEM) and atom probe (AP) could not detect any evidence of the matrix damage. In this study, post irradiation annealing experiments combining positron annihilation lineshape analysis (PALA) and hardness experiments were applied to an actual surveillance test specimen and a sample of reactor vessel steel irradiated in a material test reactor (MTR), in order to investigate the matrix damage recovery behavior and its contribution to hardening. It was confirmed that higher fluence increased the hardness and the volume fraction of open volume defects and that higher flux decreased the thermal stability of matrix damage and the effect on hardening. The contribution of matrix damage to hardening could be estimated to be below 30%. (author)

  11. Status of High Flux Isotope Reactor (HFIR) post-restart safety analysis and documentation upgrades

    International Nuclear Information System (INIS)

    Cook, D.H.; Radcliff, T.D.; Rothrock, R.B.; Schreiber, R.E.

    1990-01-01

    The High Flux Isotope Reactor (HFIR), an experimental reactor located at the Oak Ridge National Laboratory (ORNL) and operated for the US Department of Energy by Martin Marietta Energy Systems, was shut down in November, 1986 after the discovery of unexpected neutron embrittlement of the reactor vessel. The reactor was restarted in April, 1989, following an extensive review by DOE and ORNL of the HFIR design, safety, operation, maintenance and management, and the implementation of several upgrades to HFIR safety-related hardware, analyses, documents and procedures. This included establishing new operating conditions to provide added margin against pressure vessel failure, as well as the addition, or upgrading, of specific safety-related hardware. This paper summarizes the status of some of the follow-on (post-restart) activities which are currently in progress, and which will result in a comprehensive set of safety analyses and documentation for the HFIR, comparable with current practice in commercial nuclear power plants. 8 refs

  12. Post retention and post/core shear bond strength of four post systems.

    Science.gov (United States)

    Stockton, L W; Williams, P T; Clarke, C T

    2000-01-01

    As clinicians we continue to search for a post system which will give us maximum retention while maximizing resistance to root fracture. The introduction of several new post systems, with claims of high retentive and resistance to root fracture values, require that independent studies be performed to evaluate these claims. This study tested the tensile and shear dislodgment forces of four post designs that were luted into roots 10 mm apical of the CEJ. The Para Post Plus (P1) is a parallel-sided, passive design; the Para Post XT (P2) is a combination active/passive design; the Flexi-Post (F1) and the Flexi-Flange (F2) are active post designs. All systems tested were stainless steel. This study compared the test results of the four post designs for tensile and shear dislodgment. All mounted samples were loaded in tension until failure occurred. The tensile load was applied parallel to the long axis of the root, while the shear load was applied at 450 to the long axis of the root. The Flexi-Post (F1) was significantly different from the other three in the tensile test, however, the Para Post XT (P2) was significantly different to the other three in the shear test and had a better probability for survival in the Kaplan-Meier survival function test. Based on the results of this study, our recommendation is for the Para Post XT (P2).

  13. Mitigating IASCC of Reactor Core Internals by Post-Irradiation Annealing

    International Nuclear Information System (INIS)

    Was, Gary

    2017-01-01

    This final report summarizes research performed during the period between September 2012 and December 2016, with the objective of establishing the effectiveness of post-irradiation annealing (PIA) as an advanced mitigation strategy for irradiation-assisted stress corrosion cracking (IASCC). This was completed by using irradiated 304SS control blade material to conduct crack initiation and crack growth rate (CGR) experiments in simulated BWR environment. The mechanism by which PIA affects IASCC susceptibility will also be verified. The success of this project will provide a foundation for the use of PIA as a mitigation strategy for core internal components in commercial reactors.

  14. Mitigating IASCC of Reactor Core Internals by Post-Irradiation Annealing

    Energy Technology Data Exchange (ETDEWEB)

    Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States)

    2017-06-02

    This final report summarizes research performed during the period between September 2012 and December 2016, with the objective of establishing the effectiveness of post-irradiation annealing (PIA) as an advanced mitigation strategy for irradiation-assisted stress corrosion cracking (IASCC). This was completed by using irradiated 304SS control blade material to conduct crack initiation and crack growth rate (CGR) experiments in simulated BWR environment. The mechanism by which PIA affects IASCC susceptibility will also be verified. The success of this project will provide a foundation for the use of PIA as a mitigation strategy for core internal components in commercial reactors.

  15. CONSTRUCCIÓN DE UN REACTOR DISCONTINUO PARA LA OBTENCIÓN DE BIODIESEL A PARTIR DEL ACEITE DE Ricinus communis

    Directory of Open Access Journals (Sweden)

    Yolimar Fernández

    2014-01-01

    Full Text Available Se construyó un reactor discontinuo para obtener biodiesel a partir de 5 litros de extracto obtenido de la semilla de Ricinus communis. El reactor es de acero inoxidable, con longitud de 29 cm; diámetro interno de 15,24 cm y fondo cónico de 20cm de largo, espesor de la pared de 0,2cm, resistencia tubular de 1000 W y motor de 110 volt. Se extrajo y se comparó con las normas respectivas las propiedades físicas y químicas del aceite crudo. Se realizaron pruebas preliminares de transesterificación del aceite catalizadas con NaOH para constatar la viabilidad de la reacción y definir las condiciones operacionales. El biodiesel obtenido fue caracterizado y comparado con referencias presentes en la literatura. Los resultaron mostraron que es posible obtener el biocombustible en el reactor discontinuo con un grado de conversión 88%; confirmando su aplicación en reacciones de transesterificación en medio básico.

  16. Management of spent fuel from research and prototype power reactors and residues from post-irradiation examination of fuel

    International Nuclear Information System (INIS)

    1989-09-01

    The safe and economic management of spent fuel is important for all countries which have nuclear research or power reactors. It involves all aspects of the handling, transportation, storage, conditioning and reprocessing or final disposal of the spent fuel. In the case of spent fuel management from power reactors the shortage of available reprocessing capacity and the rising economic interest in the direct disposal of spent fuel have led to an increasing interest in the long term storage and management of spent fuel. The IAEA has played a major role in coordinating the national activities of the Member States in this area. It was against this background that the Technical Committee Meeting on ''Safe Management of Spent Fuel From Research Reactors, Prototype Power Reactors and Fuel From Commercial Power Reactors That Has Been Subjected to PIE (Post Irradiated Examination)'' (28th November - 1st December 1988) was organised. The aims of the current meeting have been to: 1. Review the state-of-the-art in the field of management of spent fuel from research and prototype power reactors, as well as the residues from post irradiation examination of commercial power reactor fuel. The emphasis was to be on the safe handling, conditioning, transportation, storage and/or disposal of the spent fuel during operation and final decommissioning of the reactors. Information was sought on design details, including shielding, criticality and radionuclide release prevention, heat removal, automation and remote control, planning and staff training; licensing and operational practices during each of the phases of spent fuel management. 2. Identify areas where additional research and development are needed. 3. Recommend areas for future international cooperation in this field. Refs, figs and tabs

  17. Post-accident cleanup and decommissioning of a reference pressurized-water reactor

    International Nuclear Information System (INIS)

    Murphy, E.S.; Holter, G.M.

    1982-10-01

    This paper summarizes the results of a conceptual study to evaluate the technical requirements, costs, and safety impacts of the cleanup and decommissioning of a large pressurized water reactor (PWR) involved in an accident. The costs and occupational doses for post-accident cleanup and dcommissioning are estimated to be substantially higher than those for decommissioning following the orderly shutdown of a reactor. A major factor in these cost and occupational dose increases is the high radiation environment that exists in the containment building following an accident which restricts worker access and increases the difficulty of performing certain tasks. Other factors which influence accident cleanup and decommissioning costs are requirements for the design and construction of special tools and equipment, increased requirements for regulatory approvals, and special waste management needs. Radiation doses to the public from routine accident cleanup and decommissioning operations are estimated to be below permissible radiation dose levels in unrestricted areas and within the range of annual doses from normal background

  18. Performances and microbial features of an aerobic packed-bed biofilm reactor developed to post-treat an olive mill effluent from an anaerobic GAC reactor

    Directory of Open Access Journals (Sweden)

    Marchetti Leonardo

    2006-04-01

    Full Text Available Abstract Background Olive mill wastewater (OMW is the aqueous effluent of olive oil producing processes. Given its high COD and content of phenols, it has to be decontaminated before being discharged. Anaerobic digestion is one of the most promising treatment process for such an effluent, as it combines high decontamination efficiency with methane production. The large scale anaerobic digestion of OMWs is normally conducted in dispersed-growth reactors, where however are generally achieved unsatisfactory COD removal and methane production yields. The possibility of intensifying the performance of the process using a packed bed biofilm reactor, as anaerobic treatment alternative, was demonstrated. Even in this case, however, a post-treatment step is required to further reduce the COD. In this work, a biological post-treatment, consisting of an aerobic biological "Manville" silica bead-packed bed aerobic reactor, was developed, tested for its ability to complete COD removal from the anaerobic digestion effluents, and characterized biologically through molecular tools. Results The aerobic post-treatment was assessed through a 2 month-continuous feeding with the digested effluent at 50.42 and 2.04 gl-1day-1 of COD and phenol loading rates, respectively. It was found to be a stable process, able to remove 24 and 39% of such organic loads, respectively, and to account for 1/4 of the overall decontamination efficiency displayed by the anaerobic-aerobic integrated system when fed with an amended OMW at 31.74 and 1.70 gl-1day-1 of COD and phenol loading rates, respectively. Analysis of 16S rRNA gene sequences of biomass samples from the aerobic reactor biofilm revealed that it was colonized by Rhodobacterales, Bacteroidales, Pseudomonadales, Enterobacteriales, Rhodocyclales and genera incertae sedis TM7. Some taxons occurring in the influent were not detected in the biofilm, whereas others, such as Paracoccus, Pseudomonas, Acinetobacter and Enterobacter

  19. Effect of application rates and media types on nitrogen and surfactant removal in trickling filters applied to the post-treatment of effluents from UASB reactors

    International Nuclear Information System (INIS)

    Almeida, P. G. S. de; Taveres, F. v. F.; Chernicharo, C. A. I.

    2009-01-01

    Tricking filters are a very promising alternative for the post treatment of effluents from UASB reactors treating domestic sewage,especially in developing countries. Although a fair amount of information is already available regarding organic mater removal in this combined system, very little is known in relation to nitrogen and surfactant removal in trickling filters post-UASB reactors. Therefore, the purpose of this study was to evaluate and compare the effect evaluate and compare the effect of different application rates and packing media types on trickling filters applied to the post-treatment of effluents from UASB reactors, regarding the removal of ammonia nitrogen and surfactants. (Author)

  20. Effect of application rates and media types on nitrogen and surfactant removal in trickling filters applied to the post-treatment of effluents from UASB reactors

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, P. G. S. de; Taveres, F. v. F.; Chernicharo, C. A. I.

    2009-07-01

    Tricking filters are a very promising alternative for the post treatment of effluents from UASB reactors treating domestic sewage,especially in developing countries. Although a fair amount of information is already available regarding organic mater removal in this combined system, very little is known in relation to nitrogen and surfactant removal in trickling filters post-UASB reactors. Therefore, the purpose of this study was to evaluate and compare the effect evaluate and compare the effect of different application rates and packing media types on trickling filters applied to the post-treatment of effluents from UASB reactors, regarding the removal of ammonia nitrogen and surfactants. (Author)

  1. RA-0 reactor. New neutronic calculations; Reactor RA-0. Nuevos calculos neutronicos

    Energy Technology Data Exchange (ETDEWEB)

    Rumis, D; Leszczynski, F

    1991-12-31

    An updating of the neutronic calculations performed at the RA-0 reactor, located at the Natural, Physical and Exact Sciences Faculty of Cordoba National University, are herein described. The techniques used for the calculation of a reactor like the RA-0 allows prediction in detail of the flux behaviour in the core`s interior and in the reflector, which will be helpful for experiments design. In particular, the use of WIMSD4 code to make calculations on the reactor implies a novelty in the possible applications of this code to solve the problems that arise in practice. (Author). [Espanol] En este trabajo se actualizan los calculos neutronicos realizados para el reactor RA-0, instalado en la Facultad de Ciencias Exactas, Fisicas y Naturales de la Universidad Nacional de Cordoba. Se describen los calculos realizados hasta la fecha y los resultados obtenidos. Las tecnicas incorporadas al calculo de un reactor como el RA-0 permiten predecir en detalle el comportamiento del flujo en el interior del nucleo y en el reflector, lo que sera una importante ayuda en el diseno de experimentos. En particular, el empleo del codigo WIMSD4 para calculos del reactor completo constituye una novedad en las posibles aplicaciones de ese codigo para resolver problemas que se presentan en la practica. (Autor).

  2. Methodology for the integral comparison of nuclear reactors: selecting a reactor for Mexico; Metodologia para la comparacion integral de reactores nucleares: seleccion de un reactor para Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Reyes R, R.; Martin del Campo M, C. [UNAM, Facultad de Ingenieria, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: ricarera@yahoo.com.mx

    2006-07-01

    In this work it was built a methodology to compare nuclear reactors of third generation that can be contemplated for future electric planning in Mexico. This methodology understands the selection of the reactors to evaluate eliminating the reactors that still are not sufficiently mature at this time of the study. A general description of each reactor together with their main ones characteristic is made. It was carried out a study for to select the group of parameters that can serve as evaluation indicators, which are the characteristics of the reactors with specific values for each considered technology, and it was elaborated an evaluation matrix indicators including the reactors in the columns and those indicators in the lines. Since that none reactor is the best in all the indicators were necessary to use a methodology for multi criteria taking decisions that we are presented. It was used the 'Fuzzy Logic' technique, the which is based in those denominated diffuse groups and in a system of diffuse inference based on heuristic rules in the way 'If Then consequence> ', where the linguistic values of the condition and of the consequence is defined by diffuse groups, it is as well as the rules always they transform a diffuse group into another. Later on they combine all the diffuse outputs to create a single output and an inverse transformation is made that it transfers the output from the diffuse domain to the real one. They were carried out two studies one with the entirety of the indicators and another in which the indicators were classified in three approaches: the first one refers to indicators related with the planning of the plants inside the context of the general electric sector, the second approach includes indicators related with the characteristics of the fuel and the third it considers indicators related with the acting of the plant in safety and environmental impact. This second study allowed us to know the qualities of

  3. Commissioning of research reactors. Safety guide

    International Nuclear Information System (INIS)

    2006-01-01

    The objective of this Safety Guide is to provide recommendations on meeting the requirements for the commissioning of research reactors on the basis of international best practices. Specifically, it provides recommendations on fulfilling the requirements established in paras 6.44 and 7.42-7.50 of International Atomic Energy Agency, Safety of Research Reactors, IAEA Safety Standards Series No. NS-R-4, IAEA, Vienna (2005) and guidance and specific and consequential recommendations relating to the recommendations presented in paras 615-621 of International Atomic Energy Agency, Safety in the Utilization and Modification of Research Reactors, Safety Series No. 35-G2, IAEA, Vienna (1994) and paras 228-229 of International Atomic Energy Agency, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, Safety Series No. 35-G1, IAEA, Vienna (1994). This Safety Guide is intended for use by all organizations involved in commissioning for a research reactor, including the operating organization, the regulatory body and other organizations involved in the research reactor project

  4. Post-accident cleanup and decommissioning of a reference pressurized water reactor

    International Nuclear Information System (INIS)

    Murphy, E.S.; Holter, G.M.

    1982-01-01

    This paper summarizes the results of a conceptual study to evaluate the technical requirements, costs, and safety impacts of the cleanup and decommissioning of a large pressurized water reactor (PWR) involved in an accident. The costs and occupational doses for post-accident cleanup and decommissioning are estimated to be substantially higher than those for decommissioning following the orderly shutdown of a reactor. A major factor in these cost and occupational dose increases is the high radiation environment that exists in the containment building following an accident which restricts worker access and increases the difficulty of performing certain tasks. Other factors which influence accident cleanup and decommissioning costs are requirements for the design and construction of special tools and equipment, increased requirements for regulatory approvals, and special waste management needs. Radiation doses to the public from routine accident cleanup and decommissioning operations are estimated to be below permissible radiation dose levels in unrestricted areas and within the range of annual doses from normal background. 6 references, 1 figure, 7 tables

  5. Bioreduction of para-chloronitrobenzene in drinking water using a continuous stirred hydrogen-based hollow fiber membrane biofilm reactor

    Energy Technology Data Exchange (ETDEWEB)

    Xia Siqing, E-mail: siqingxia@gmail.com [State Key Laboratory of Pollution Control and Resource Reuse, College of Environmental Science and Engineering, Tongji University, Shanghai 200092 (China); Li Haixiang; Zhang Zhiqiang [State Key Laboratory of Pollution Control and Resource Reuse, College of Environmental Science and Engineering, Tongji University, Shanghai 200092 (China); Zhang Yanhao [College of Municipal and Environmental Engineering, Shandong Jianzhu University, Jinan 250101 (China); Yang Xin; Jia Renyong; Xie Kang; Xu Xiaotian [State Key Laboratory of Pollution Control and Resource Reuse, College of Environmental Science and Engineering, Tongji University, Shanghai 200092 (China)

    2011-08-30

    Highlights: {yields} We designed a novel hollow fiber membrane biofilm reactor for p-CNB removal. {yields} Biotransformation pathway of p-CNB in the reactor was investigated in this study. {yields} Nitrate and sulfate competed more strongly for hydrogen than p-CNB. {yields} This reactor achieved high removal efficiency and hydrogen utilization efficiency. - Abstract: para-Chloronitrobenzene (p-CNB) is particularly harmful and persistent in the environment and is one of the priority pollutants. A feasible degradation pathway for p-CNB is bioreduction under anaerobic conditions. Bioreduction of p-CNB using a hydrogen-based hollow fiber membrane biofilm reactor (HFMBfR) was investigated in the present study. The experiment results revealed that p-CNB was firstly reduced to para-chloraniline (p-CAN) as an intermediate and then reduced to aniline that involves nitro reduction and reductive dechlorination with H{sub 2} as the electron donor. The HFMBfR had reduced p-CNB to a major extent with a maximum removal percentage of 99.3% at an influent p-CNB concentration of 2 mg/L and a hydraulic residence time of 4.8 h, which corresponded to a p-CNB flux of 0.058 g/m{sup 2} d. The H{sub 2} availability, p-CNB loading, and the presence of competing electron acceptors affected the p-CNB reduction. Flux analysis indicated that the reduction of p-CNB and p-CAN could consume fewer electrons than that of nitrate and sulfate. The HFMBfR had high average hydrogen utilization efficiencies at different steady states in this experiment, with a maximum efficiency at 98.2%.

  6. Experimental framework for laboratory scale microgrids

    Directory of Open Access Journals (Sweden)

    José Alex Restrepo-Zambrano

    2016-01-01

    Full Text Available Este artículo presenta una propuesta de un banco de pruebas de microrredes para uso en laboratorio. El objetivo es proporcionar alta flexibilidad utilizando un enfoque modular con un hardware común para la mayoría de las tareas. El marco experimental propuesto para microrredes a escala de laboratorio proporciona los requisitos para enseñanza e investigación. Esto se logra con una etapa de electrónica de potencia reconfigurable, para pruebas y diseños de nuevas topologías. Permite probar algoritmos en los distintos niveles de la estructura jerárquica de la microrred. Da acceso a la emulación y simulación de elementos encontrados comúnmente en una microrred y a la programación de bajo nivel de los protocolos de comunicación para estudiar el canal de comunicación. La unidad de procesamiento en cada módulo, llamado controlador local, utiliza un procesador digital de señales de alto rendimiento (DSP. Esta unidad de procesamiento permite la reconfiguración de cada módulo para asumir cualquier tarea en la microrred; es decir, como cargas controlables, almacenamiento de energía, generación eólica, generación fotovoltaica, etc. El hardware propuesto se probó operando como emulador de los diferentes subsistemas. Las comunicaciones con un controlador central microrred (MCC se realizan mediante procesadores integrados estándar, capaces de implementar los protocolos de comunicación adecuados para ambientes de microrred.

  7. Análisis de estabilidad del reactor PFTR para una reacción con cinética de primer orden utilizando la funcional de Lyapunov

    Directory of Open Access Journals (Sweden)

    Héctor Armando Durán Peralta

    2007-01-01

    Full Text Available Abunda la literatura referente al análisis de estabilidad de reactores con parámetros globalizados de concentración y temperatura (por ejemplo el CSTR, en cambio es escasa la literatura sobre la estabilidad de reactores con parámetros distribuidos donde existe distribución espacial de concentración y temperatura, como es el caso del reactor tubular PFTR. Este documento analiza la estabilidad del reactor PFTR isotérmico y no isotérmico para una reacción con cinética de primer orden utilizando la funcional de Lyapunov. Se trabaja con una cinética de primer orden pues un objetivo de este artículo es mostrar cómo se aplica la funcional de Lyapunov al análisis de un reactor de parámetros distribuidos, dado que es casi inexistente la literatura sobre el método de la funcional de Lyapunov aplicada a la estabilidad de reactores (técnica usada en el análisis de estabilidad de sistemas en ingeniería eléctrica. El análisis de estabilidad dio como resultado perfiles de temperatura y concentración asintóticamente estables para los casos PFTR isotérmico, no isotérmico con constante cinética independiente de la temperatura y PFTR no isotérmico adiabático. Para el PFTR con retiro de calor el análisis condujo a una región de estabilidad asintótica y a una región incierta donde puede o no haber oscilaciones.

  8. Dispositivo de posicionamiento de muestras biológicas para su irradiación en un canal radial de un reactor nuclear // Biological samples positioning device for irradiations on a radial channel at the nuclear research reactor

    Directory of Open Access Journals (Sweden)

    Maritza Rodríguez - Gual

    2010-05-01

    Full Text Available ResumenPor la demanda de un dispositivo experimental para el posicionamiento de las muestras biológicaspara su irradiación en un canal radial de un reactor nuclear de investigaciones en funcionamiento, seconstruyó y se puso en marcha un dispositivo para la colocación y retirada de las muestras en laposición de irradiación de dicho canal. Se efectuaron las valoraciones económicas comparando conotro tipo de dispositivo con las mismas funciones. Este trabajo formó parte de un proyectointernacional entre Cuba y Brasil que abarcó el estudio de los daños inducidos por diferentes tipos deradiación ionizante en moléculas de ADN. La solución propuesta es comprobada experimentalmente,lo que demuestra la validez práctica del dispositivo. Como resultado del trabajo, el dispositivoexperimental para la irradiación de las muestras biológicas se encuentra instalado y funcionando yapor 5 años en el canal radial # 3(BH#3 Palabras claves: reactor nuclear de investigaciones, dispositivo para posicionamiento de muestras,___________________________________________________________________________AbstractFor the demand of an experimental device for biological samples positioning system for irradiationson a radial channel at the nuclear research reactor in operation was constructed and started up adevice for the place and remove of the biological samples from the irradiation channels withoutinterrupting the operation of the reactor. The economical valuations are effected comparing withanother type of device with the same functions. This work formed part of an international projectbetween Cuba and Brazil that undertook the study of the induced damages by various types ofionizing radiation in DNA molecules. Was experimentally tested the proposed solution, whichdemonstrates the practical validity of the device. As a result of the work, the experimental device forbiological samples irradiations are installed and operating in the radial beam hole #3(BH#3

  9. Nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

    This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

  10. Underground collocation of nuclear power plant reactors and repository to facilitate the post-renaissance expansion of nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Myers, Carl W [Los Alamos National Laboratory; Elkins, Ned Z [Los Alamos National Laboratory

    2008-01-01

    Underground collocation of nuclear power reactors and the nuclear waste management facilities supporting those reactors, termed an underground nuclear park (UNP), appears to have several advantages compared to the conventional approach to siting reactors and waste management facilities. These advantages include the potential to lower reactor capital and operating cost, lower nuclear waste management cost, and increase margins of physical security and safety. Envirorunental impacts related to worker health, facility accidents, waste transportation, and sabotage and terrorism appear to be lower for UNPs compared to the current approach. In-place decommissioning ofUNP reactors appears to have cost, safety, envirorunental and waste disposal advantages. The UNP approach has the potential to lead to greater public acceptance for the deployment of new power reactors. Use of the UNP during the post-nuclear renaissance time frame has the potential to enable a greater expansion of U.S. nuclear power generation than might otherwise result. Technical and economic aspects of the UNP concept need more study to determine the viability of the concept.

  11. Reactor network synthesis for isothermal conditions = Síntese de redes de reatores para condições isotérmicas

    Directory of Open Access Journals (Sweden)

    Lincoln Kotsuka da Silva

    2008-07-01

    Full Text Available In the present paper, a computational systematic procedure for isothermal Reactor Network Synthesis (RNS is presented. A superstructure of ideal CSTR and PFR reactors is proposed and the model is formulated as a constrained Nonlinear Programming (NLP problem. Complex reactions (series/parallel reactions are considered. The objective function is based on yield or selectivity, depending on the desired product, subject to different operational conditions. The problem constraints are mass balances in the reactorsand in the considered reactor network superstructure. A systematic computational procedure is proposed and a Genetic Algorithm (GA is developed to obtain the optimal reactor arrangement with the maximum yield or selectivity and minimum reactor volume. Results are as good as or better than those reported in the literature.No presentetrabalho apresenta-se um procedimento computacional para síntese de redes de reatores (SRR operando em condições isotérmicas. Uma superestrutura de rede de reatores formada por reatores ideais CSTR e PFR é proposta e o problema apresenta uma formulação de programação não linear (PNL. São consideradas reações complexas (série/paralelas. A função objetivo é baseada no rendimento ou na seletividade em relação ao produto desejado, sujeito a diferentes condições de operação. As restrições ao problema são provenientes dos balanços de massa e da configuração da superestrutura considerada.No procedimento computacional é proposto um Algoritmo Genético (AG para obtenção do arranjo ótimo de reatores com máximo rendimento ou seletividade com menor volume reacional. Os resultados obtidos são condizentes com os obtidos na literatura.

  12. Análisis de estabilidad del reactor PFTR para una reacción con cinética de primer orden utilizando la funcional de Lyapunov

    OpenAIRE

    Héctor Armando Durán Peralta; Luis Fernando Córdoba C

    2007-01-01

    Abunda la literatura referente al análisis de estabilidad de reactores con parámetros globalizados de concentración y temperatura (por ejemplo el CSTR), en cambio es escasa la literatura sobre la estabilidad de reactores con parámetros distribuidos donde existe distribución espacial de concentración y temperatura, como es el caso del reactor tubular PFTR. Este documento analiza la estabilidad del reactor PFTR isotérmico y no isotérmico para una reacción con cinética de primer orden utilizando...

  13. Análisis de estabilidad del reactor pftr para una reacción con cinética de primer orden utilizando la funcional de lyapunov

    OpenAIRE

    Durán Peralta, Héctor Armando; Córdoba C, Luis Fernando

    2010-01-01

    Abunda la literatura referente al análisis de estabilidad de reactores con parámetros globalizados de concentración y temperatura (por ejemplo el CSTR), en cambio es escasa la literatura sobre la estabilidad de reactores con parámetros distribuidos donde existe distribución espacial de concentración y temperatura, como es el caso del reactor tubular PFTR. Este documento analiza la estabilidad del reactor PFTR isotérmico y no isotérmico para una reacción con cinética de primer orden utilizand...

  14. Postes prefabricados de fibra: Consideraciones para su uso clínico

    Directory of Open Access Journals (Sweden)

    Hugo Calabria Díaz

    Full Text Available En la década de los 90 los Postes Prefabricados de Fibra (PPF se introdujeron al mercado como alternativa a los sistemas metálicos o cerámicos. Hasta la fecha se mantienen en uso, modificando de manera constante sus presentaciones comerciales y estrategias de fijación. Sus cualidades mecánicas como su bajo Módulo Elástico (ME similar al dentinario, introducen un nuevo paradigma en la rehabilitación del Diente Endodonticamente Tratado (DET: .ldquo;el poste debe acompañar en forma solidaria la flexión de los tejidos dentarios frente a las cargas.rdquo;. Sus actuales cualidades estéticas, la fácil remoción y la posibilidad de su cementado adhesivo, los han convertido en una alternativa válida a las soluciones convencionales. Sin embargo, algunos resultados contradictorios junto con la importante dificultad de lograr hibridación en la dentina radicular, mantienen interrogantes a resolver en el futuro. Se indican en casos en donde se prevea retratamiento, en pacientes jóvenes, de alta exigencia estética y toda vez que se quiera y pueda eludir los costos de aleaciones nobles. En el presente artículo se analizan los fundamentos clínicos y experimentales de distintos autores, extrapolándose consideraciones prácticas para su uso. Los mismos se ilustran en un caso clínico para un paciente joven, con antecedente de traumatismo y con altas expectativas estéticas

  15. Observation and control system of the thermohydraulic assays laboratory; Sistema de observacion y control del laboratorio de ensayos termohidraulicos

    Energy Technology Data Exchange (ETDEWEB)

    Santome, D; Hualde, R

    1991-12-31

    The Thermohydraulic Assays Laboratory (L.E.T.) is an installation whose purpose will be the components testing and the CAREM-25 reactor thermohydraulic processes operation dynamics. This plant is located at Pilcaniyeu, province of Rio Negro. Part of the tests which will be carried out consist in the use of different control strategies. The control of the systems by digital processors (control by software) has been decided to proceed with a maximum flexibility and capacity to make changes in the algorithms. This work describes the design and implementation of a digital control system to command the three circuits of the installation. (Author). [Espanol] El Laboratorio de Ensayos Termohidraulicos (L.E.T.) es una instalacion cuyo objeto sera el ensayo de componentes y de la dinamica de operacion de los procesos termohidraulicos del reactor CAREM-25. Esta planta esta localizada en Pilcaniyeu, provincia de Rio Negro. Parte de las pruebas que se efectuaran en el L.E.T. consisten en el empleo de distintas estrategias de control. Para disponer de una maxima flexibilidad y capacidad de efectuar cambios en los algoritmos, se decidio realizar el control de los sistemas por medio de procesadores digitales (control por software). Este trabajo consistio en el diseno e implementacion de un sistema de control digital distribuido para el comando de los tres circuitos con que cuenta la instalacion. (Autor).

  16. The post-irradiated examination of CANDU type fuel irradiated in the Institute for Nuclear Research TRIGA reactor

    International Nuclear Information System (INIS)

    Tuturici, I.L.; Parvan, M.; Dobrin, R.; Popov, M.; Radulescu, R.; Toma, V.

    1995-01-01

    This post-irradiation examination work has been done under the Research Contract No. 7756/RB, concluded between the International Atomic Energy Agency and the Institute for Nuclear Research. The paper contains a general description of the INR post-irradiation facility and methods and the relevant post-irradiation examination results obtained from an irradiated experimental CANDU type fuel element designed, manufactured and tested by INR in a power ramp test in the 100 kW Pressurised Water Irradiation Loop of the TRIGA 14 MW(th) Reactor. The irradiation experiment consisted in testing an assembly of six fuel elements, designed to reach a bumup of ∼ 200 MWh/kgU, with typical CANDU linear power and ramp rate. (author)

  17. Post-accident monitoring systems in Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Suriya Murthy, N.; Sivasailanathan, Vidhya; Ananth, Allu; Roy, Kallol

    2018-01-01

    PFBR is a 500 MW(e) MOX fueled and sodium cooled fast reactor (SFR) under advanced stage of commissioning at Kalpakkam. Currently, the main vessel is preheated and sodium has been charged into two secondary loops that are operated in recirculation mode. In order to characterize the radiation field and contamination, the workplace monitoring is undertaken using installed monitors that are commissioned and made operational. This helps to ensure radiological protection during normal operating conditions. On the other hand, radiological monitoring in emergency conditions is quite different. For undertaking the mitigative accident management, a set of specialized nuclear instruments called post-accident monitoring systems (PAMS) which include radiation monitors are stipulated. The Fukushima Daiichi accident emphasized the importance and need for reliable accident monitoring instrumentation to indicate the safety functions during the progression and aftermath of accident in NPP. In PFBR, the PAMS are integrated with other monitoring systems in design stage itself to manage the measurements and indicating the safety functions for implementing EOP and SAMG

  18. Post-treatment of Fly Ash by Ozone in a Fixed Bed Reactor

    DEFF Research Database (Denmark)

    Pedersen, Kim Hougaard; Melia, M. C.; Jensen, Anker Degn

    2009-01-01

    to be fast. A kinetic model has been formulated, describing the passivation of carbon, and it includes the stoichiometry of the ozone consumption (0.8 mol of O-3/kg of C) and an ineffective ozone loss caused by catalytic decomposition. The simulated results correlated well with the experimental data....... prevents the AEA to be adsorbed. In the present work, two fly ashes have been ozonated in a fixed bed reactor and the results showed that ozonation is a potential post-treatment method that can lower the AEA requirements of a fly ash up to 6 times. The kinetics of the carbon oxidation by ozone was found...

  19. Approach for estimating post-annual reirradiation embrittlement of reactor vessel steels

    International Nuclear Information System (INIS)

    Server, W.L.; Taboada, A.

    1985-01-01

    Thermal annealing of a commercial nuclear reactor pressure vessel is a possible solution for extending lifetime in situations where excessive radiation embrittlement has taken place or when the original design life is approached. Two difficult facets of thermal annealing are the degree of toughness recovery after annealing and the post-anneal reirradiation embrittlement behavior. These aspects of annealing are evaluated in this paper by using simple models and translation of the initial irradiation damage curve either vertically or laterally at the point of residual shift after annealing. Results using this methodology are compared to limited actual weld metal measurements of annealing behavior. A forthcoming ASTM Guide on in-place annealing uses this methodology to assess annealing recovery and re-embrittlement response

  20. Análisis y modelado del rendimiento de algoritmos paralelos en clusters de computadoras

    OpenAIRE

    Rivera Zamarripa, Luis Alberto

    2016-01-01

    Hoy en día los clusters junto con las bibliotecas de paso de mensajes (tal como MPI)son una buena alternativa para ejecutar aplicaciones paralelas que requieren de mucho poder de cómputo. Esto es debido a su bajo costo y su cada vez mejor desempeño. Cuando una aplicación paralela es ejecutada en un cluster su código y los datos son distribuidos entre los procesadores para su procesamiento con el fin de obtener un buen desempeño. Sin embargo, para realizar esta distribución y obten...

  1. Síntesis y evaluación de un DSP empotrado en una FPGA

    Directory of Open Access Journals (Sweden)

    Juan Raúl Rodríguez Suárez

    2010-07-01

    Full Text Available Normal 0 21 false false false ES-TRAD X-NONE X-NONE MicrosoftInternetExplorer4 En el trabajo se utiliza el procesador MicroBlaze, como módulo IP (Intellectual Property empotrado en la tarjeta Spartan-3E de la compañía Xilinx, para realizar aplicaciones de Procesamiento Digital de Señales (PDS tales como filtros del tipo FIR (Finite Impulse Response, IIR (Infinite Impulse Response, y de análisis espectral como la FFT (Fast Fourier Transform; debido a que esta tarjeta no posee embebido un procesador digital. Para ello se genera una señal a través del software Matlab, es enviada hacia la tarjeta Spartan-3E, se procesa con una de las aplicaciones, y los resultados son devueltos a Matlab, donde son comprobados mediante la comparación con los resultados de las funciones existentes en dicho software relativas a las aplicaciones dadas. La comunicación entre la PC (Personal Computer y la tarjeta se hace a través del puerto serie. Los resultados de todas las aplicaciones producidas con MicroBlaze coinciden con los de Matlab

  2. POST: a postprocessor computer code for producing three-dimensional movies of two-phase flow in a reactor vessel

    International Nuclear Information System (INIS)

    Taggart, K.A.; Liles, D.R.

    1977-08-01

    The development of the TRAC computer code for analysis of LOCAs in light-water reactors involves the use of a three-dimensional (r-theta-z), two-fluid hydrodynamics model to describe the two-phase flow of steam and water through the reactor vessel. One of the major problems involved in interpreting results from this code is the presentation of three-dimensional flow patterns. The purpose of the report is to present a partial solution to this data display problem. A first version of a code which produces three-dimensional movies of flow in the reactor vessel has been written and debugged. This code (POST) is used as a postprocessor in conjunction with a stand alone three-dimensional two-phase hydrodynamics code (CYLTF) which is a test bed for the three-dimensional algorithms to be used in TRAC

  3. Methodology for monitoring the behaviour of wind-photovoltaic hybrid system in the conditions of Cuba; Metodologia para el monitoreo del comportamiento de un sistema hibrido eolico-fotovoltaico en las condiciones de Cuba

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez G, Maria; Nunez, Ariel; Marquez M, Soe del C [Centro de Investigaciones de Energia Solar, Santiago de Cuba (Cuba)

    2000-07-01

    The proposal of a methodology is shown in the work that allows monitoring the behaviour of wind-photovoltaic hybrid system beginning with the study of the energy resources (wind and solar) of the known place, designed and put into operation a hybrid installation, using Text Processing techniques could obtained operation curves of the system daily, monthly and annual, to configure the reading for port series of the parameters measured during the evaluation of the system a denominated software HYBSYS was developed in Lab View for Windows 3.1 or superior. [Spanish] Se muestra la propuesta de una metodologia que permite monitorear el comportamiento de un sistema hibrido eolico-fotovoltaico, comenzando con el estudio de los recursos energeticos (eolico y solar) de un sitio conocido se diseno y puso en funcionamiento una instalacion hibrida, usando las tecnicas de un procesador se pudieron obtener las curvas de funcionamiento diaria, mensual y anual del sistema, para configurar la lectura por puerto serie de los parametros medidos durante la evaluacion del sistema se desarrollo un software denominado HYBSYS en Lab View para Windows 3.1 o superior.

  4. Design considerations for post accident monitoring system of a research reactor

    International Nuclear Information System (INIS)

    Jang, Gwi Sook; Park, Je Yun; Kim, Young Ki

    2012-01-01

    The Post Accident Monitoring System (PAMS) provides primary information for operators to assess the plant conditions and perform their role in bringing the plant to a safe condition during an accident. The PAMS of NPP (Nuclear Power Plant) in KOREA provides the continuous display of the PAM category 1 parameters specified in R.G 1.97, Rev. 03. Recently the PAMS of NPP has been designed according to R.G 1.97, Rev. 04. There is no PAMS at the HANARO in KOREA, but recently RRs (Research Reactors) around the world are going to have PAMS for various multi purposes. We should determine the design considerations for PAMS in a Korean RR based on the design state analysis. Thus, this paper proposes strategies on the design considerations for the PAMS of a Korean RR

  5. Accionamiento de un ventilador industrial para prueba de aerogeneradores; Drive of an industrial fan for wind testing

    Directory of Open Access Journals (Sweden)

    Francisco Eneldo López Monteagudo

    2015-04-01

    Full Text Available En este trabajo se implementó el control de un ventilador industrial utilizado para prueba de aerogeneradores, el cual es empleado como un dispositivo interno en la elaboración de un túnel de viento, para realizar pruebas de medición de viento. El proyecto consistió en regular la velocidad de un ventilador industrial utilizado en un túnel de viento, para realizar pruebas de sistemas de control en aerogeneradores, generándose señales de viento reguladas en valores constantes, ó que sigan un perfil definido por una base de datos de valores reales medidos con un anemómetro. Para implementar el sistema de control y la comunicación de los dispositivos, se empleó un procesador digital de señales (PDS de Texas Instruments EZDSP2407, que actúa como interfaz para transmitir los datos entre el entorno de programación (VisSim Embedded Control Developer (ECD. Además se utilizó un variador de velocidad de 3HP de la marca SIEMENS modelo Micromaster 420. In this work, a fan control industrial wind turbines used for test, which is used as an internal device in the development of a wind tunnel for testing wind measurement. The project consists of regulating the speed of an industrial fan used in a wind tunnel to test control systems in wind turbines, wind generating regulated signals in constant, or to follow a profile defined by a database of values actual measured with an anemometer. To implement the control system and communication devices, in this project employed a digital signal processor (DSP from Texas Instruments EZDSP2407, which acts as an interface to transmit data between the programming environments (VisSim Embedded Control Developer (ECD. Also uses a variable speed 3HP SIEMENS Micromaster model 420.

  6. Patrón pipeline aplicado a arquitecturas heterogéneas big.LITTLE

    OpenAIRE

    Vilches, Antonio; Rodriguez, Andres; Navarro, Ángeles; Corbera, Francisco; Asenjo-Plaza, Rafael

    2015-01-01

    En este trabajo, proponemos una solución para permitir la ejecución de aplicaciones de tipo streaming, que constan de una serie de etapas, sobre arquitecturas heterogéneas con un multicore y una GPU integrada. Para ello, presentamos una API que permite especificar el nivel de paralelismo explotado en el multicore, la asignación de las etapas del pipeline a los procesadores (CPU y GPU), y el número de threads. Usando una aplicación real de tipo streaming como caso de estudio, evalu...

  7. Post-irradiation examination of Al-61 wt% U3Si fuel rods from the NRU reactor

    International Nuclear Information System (INIS)

    Sears, D.F.; Wang, N.

    1997-01-01

    This paper describes the post-irradiation examination of 4 intact low enrichment uranium (LEU) fuel rods from the national research universal (NRU) reactor at the Chalk River Laboratories of AECL. The rods were irradiated during the period 1993 through 1995, under typical driver fuel operating conditions in NRU, i.e., nominal D 2 O coolant inlet temperature 37E C, inlet pressure 654 kPa and mass flow 12.4 L/s. Irradiation exposures ranged from 147 to 251 full-power days, corresponding to 40 to 84 atom % 235 U burnup. The maximum rod power was ∼2 MW, with element linear power ratings up to 68 kW/m. Post-irradiation examinations, conducted in 1997, focused on optical metallography to measure cladding oxide thickness and fuel core and cladding microstructural examinations. The cladding oxide was approximately 24 : m thick at the mid-plane of fuel rods irradiated to 251 full-power days, with small areas up to 34 : m thick on the fins. The cladding retained significant ductility after irradiation, and its microstructure appeared unchanged. Fuel core diametral increases were small (up to 4%) and within the range previously observed on Al-61 wt% U 3 Si fuel irradiated in the NRU reactor. (author)

  8. Estudio de la distribución de los tiempos de residencia en un reactor tubular para la hidrólisis de lecitina de soja con fosfolipasa A2 inmovilizada

    Directory of Open Access Journals (Sweden)

    Zaritzky, N.

    2001-10-01

    efectuada mediante el uso de enzima fosfolipasa A2 inmovilizada, liberando un ácido graso de la posición C-2 de los fosfolípidos para obtener un producto enriquecido en lisolecitinas. La reacción enzimática sigue una cinética de primer orden cuando las concentraciones de sustrato están dentro del rango: 6,34 10-3 y 19,0 10-3M. El valor de la constante de velocidad es: k= 9,88 10-2 min-1 cuando la enzima está inmovilizada sobre alúmina. Se construyó un reactor que permite la circulación del fluido a través del soporte. El soporte seleccionado fue alúmina en consideración a sus buenas propiedades mecánicas y a su bajo costo. Fue analizado el comportamiento del flujo en el reactor, y cuanto este se aparta del modelo ideal de flujo en pistón, inyectando una solución de 1 % NaCl (trazador en forma de inyección por impulso. La medición de la conductividad de la solución efluente resultó adecuada para la determinación de los tiempos de residencia. El sistema mostró comportamiento lineal. Se analizaron los tiempos de residencia en el reactor utilizando tres diferentes volúmenes de flujo para diferentes arreglos de soporte y material inerte. Se calcularon las fracciones no convertidas en el reactor y se observaron las diferencias a la salida en comparación a las de un reactor de flujo en pistón, precisamente porque se generan canalizaciones y cortocircuitos en la columna. La conversión máxima resultó para las más altas concentraciones de sustrato y para el menor flujo de alimentación. El módulo de dispersión resultó bastante mayor que el límite que introduce una curva gaussiana para el caso en el cual el grado de suposición de alta dispersión fue corregido. El reactor alcanzó un comportamiento similar al de un reactor de mezcla completa y se concluyó que son importantes el grado de retromezcla, la formación de remolinos y zonas de redistribución de material.

  9. Actualización del sistema SCADA y de control para los reactores MQ5 y MQ6 de la planta de Pinturas Condor, Sherwin Williams Ecuador

    Directory of Open Access Journals (Sweden)

    Jonathan Reinoso

    2013-12-01

    Full Text Available El presente documento describe la actualización del sistema SCADA para los reactores MQ5 y MQ6 de la planta de Pinturas Condor mediante el software Intouch y la actualización del sistema de control del reactor MQ5 implementado en un controlador lógico programable (PLC de marca SCHNEIDER, además de la arquitectura de control realizada en el proyecto. El sistema SCADA y de control de los reactores permiten la visualización y control de los datos y variables más relevantes durante las diferentes fases de producción de resinas en los reactores MQ5 y MQ6.

  10. Metodología para resolver la ecuación del transporte con el código de ordenadas discretas TORT en el reactor IPEN/MB-01

    OpenAIRE

    Bernal, A.; Abarca Giménez, Agustín; Barrachina Celda, Teresa María; Miró Herrero, Rafael; Verdú Martín, Gumersindo Jesús

    2013-01-01

    La resolución de la Ecuación del Transporte Neutrónico en estado estacionario en reactores nucleares de tipo piscina, se consigue normalmente por medio de 2 métodos numéricos diferentes: Monte Carlo (estocástico) y Ordenadas Discretas (determinista). El método de las Ordenadas Discretas resuelve la Ecuación del Transporte Neutrónico para un conjunto de determinadas direcciones, obteniendo un conjunto de ecuaciones y soluciones para cada dirección, donde la solución para cada dirección es el f...

  11. Matrix multiplication with a hypercube algorithm on multi-core processor cluster

    Directory of Open Access Journals (Sweden)

    José Crispín Zavala-Díaz

    2015-01-01

    Full Text Available Se analiza, modifica e implementa el algoritmo de multiplicación de matrices de Dekel, Nassimi y Sahani o hipercubo en un cluster de procesadores multi-core, donde el número de procesadores utilizado es menor al requerido por el algoritmo de n3. Se utilizan 23, 43 y 83 unidades procesadoras para multiplicar matrices de orden de magnitud de 10X10, 102X102 y 103X103. Los resultados del modelo matemático del algoritmo modificado y los obtenidos de la experimentación computacional muestran que es posible alcanzar rapidez y eficiencias paralelas aceptables, en función del número de unidades procesadoras utilizadas. También se muestra que la influencia del enlace externo de comunicación entre los nodos disminuye si se utiliza una combinación de los canales de comunicación disponibles entre los núcleos en un cluster multi-core.

  12. CÓMPUTO DE ALTO DESEMPEÑO PARA OPERACIONES VECTORIALES EN BLAS-1 // INCREASED COMPUTATIONAL PERFORMANCE FOR VECTOR OPERATIONS ON BLAS-1

    Directory of Open Access Journals (Sweden)

    José Antonio Muñoz Gómez

    2014-06-01

    Full Text Available The functions library, called Basic Linear Algebra Subprograms (BLAS-1, is considered the programming standard in scientific computing. In this work, we focus on the analysis of various code optimization techniques to increase the computational performance of BLAS-1. In particular, we address a combinational approach to explore possible methods of encoding using unroll technique with different levels of depth, vector data programming with MMX and SSE for Intel processors. Using the main functions of BLAS-1, it was determined numerically a computational increase, expressed in mega-ops, up to 52% compared to the optimized BLAS-1 ATLASlibrary.// RESUMEN: La biblioteca de funciones denominada Subprogramas Básicos de Algebra Lineal (BLAS-1 es considerada el estándar de programación en computación científica. En este trabajo nos enfocamos en el análisis de diversas técnicas de optimización de código para incrementar el desempeño computacional de BLAS-1. En particular abordamos un enfoque combinacional para explorar las posibles formas de codificación empleando la técnica de unroll con diversos niveles de profundidad, programación vectorial de datos con MMX y SSE para procesadores Intel. Empleando las funciones principales de BLAS-1 determinamos numéricamente un incremento computacional, expresado en mega-flops, de hasta 52% en comparación con la biblioteca optimizada BLAS-1 de ATLAS.

  13. Informe sobre Ensayo de Flexión Estática en Postes de Ciprés (Cupressus Lusitánica Miller, para Instalaciones Eléctricas y Telefónicas.

    Directory of Open Access Journals (Sweden)

    Hoheisel Hannes

    1973-06-01

    Full Text Available Los postes de madera utilizados para tendido de líneas eléctricas y telefónicas están sometidos a dos tipos de esfuerzos: a Esfuerzo de compresión axial, debido al peso del poste y de los hilos. b Esfuerzo de flexión debido a la acción del viento o a una tensión accidental de los hilos. Para efectos de cálculo, la compresión axial no se toma en cuenta por ser muy pequeña; pero la flexión tiene importancia y es el ensayo básico para el diseño de postes. El ensayo de postes a la flexión se realiza en postes de tamaño natural (9-10 m. . Se considera el presente informe como preliminar, debido a que no existe en este momento una clasificación de los postes en cuanto a la resistencia a la flexión. El Laboratorio de Productos Forestales de la Universidad Nacional, Sede de Medellín, tiene en su plan de trabajo un estudio de clasificación de postes con base en la resistencia a la flexión y de acuerdo con el rango de diámetros que puedan utilizarse, para correlacionar datos de resistencia de postes de diferentes tamaños.

  14. Post-irradiation examination of A1-61 wt % U3Si fuel rods from the NRU reactor

    International Nuclear Information System (INIS)

    Sears, D.F.; Wang, N.

    1997-09-01

    This paper describes the post-irradiation examination of 4 intact low-enrichment uranium (LEU) fuel rods from the national research universal (NRU) reactor at the Chalk River Laboratories of AECL. The rods were irradiated during the period 1993 through 1995, under typical driver fuel operating conditions in NRU, i.e., nominal D 2 0 coolant inlet temperature 37 degrees C, inlet pressure 654 kPa and mass flow 12.4 L/s. Irradiation exposures ranged from 147 to 251 full-power days, corresponding to 40 to 84 atom % 235 U burnup. The maximum rod power was ∼2 MW, with element linear power ratings up to 68 kW/m. Post-irradiation examinations, conducted in 1997, focused on optical metallography to measure cladding oxide thickness and fuel core and cladding microstructural examinations. The cladding oxide was approximately 24 μm thick at the mid-plane of fuel rods irradiated to 251 full-power days, with small areas up to 34 μm thick on the fins. The cladding retained significant ductility after irradiation, and its microstructure appeared unchanged. Fuel core diametral increases were small (up to 4%) and within the range previously observed on A1-61 wt % U 3 Si fuel irradiated in the NRU reactor. (author)

  15. Pós-tratamento de efluente nitrificado da parboilização de arroz utilizando desnitrificação em reator UASB Post-treatment a nitrified parboilized rice wastewater using denitrification in UASB reactor

    Directory of Open Access Journals (Sweden)

    Loraine Andre Isoldi

    2005-12-01

    Full Text Available Um sistema combinado reator UASB-reator aeróbio foi utilizado para a remoção de nitrogênio total e DQO de efluente de parboilização de arroz. O experimento foi realizado em reatores de bancada, com volumes de 4 L (UASB e 3,6 L (reator aeróbio. Os parâmetros de operação pH, temperatura, alcalinidade e concentração de ácidos voláteis foram monitorados durante o período experimental. Para o reator aeróbio de mistura completa, foi determinada, também, a concentração de oxigênio dissolvido. O sistema combinado reator UASB-reator aeróbio apresentou uma eficiência de remoção de carbono de 84% e uma eficiência de remoção de nitrogênio total Kjeldahl de 83%. O sistema proposto, nas condições experimentais, demonstrou ser adequado para remoção, simultânea, de DQO e de compostos oxidados de nitrogênio, em reator UASB.An UASB-aerobic reactor system was used for the removal of total nitrogen and COD of effluent from industries of parboilized rice. The experiment was performed in reactors with volumes of 4 L (UASB reactor and 3,6 L (aerobic reactor, respectevely. Temperature, pH, alkalinity and volatile acids concentration were monitored during the experiment. Dissolved oxygen concentration was determined for the aerobic reactor. The UASB-aerobic reactor system showed 84% carbon removal efficiency and 83% total Kjeldahl nitrogen removal efficiency. This system was able to remove, efficiently, COD and nitrogen in an UASB reactor.

  16. SISTEMA COMBINADO DE FLOTAÇÃO POR AR DISSOLVIDO E FILTRAÇÃO ADSORTIVA EM ZEÓLITA PARA TRATAMENTO DE EFLUENTE DE REATOR UASB

    Directory of Open Access Journals (Sweden)

    Bruno Oliveira Freitas

    2016-06-01

    Full Text Available RESUMO: Dentre as soluções existentes para efetuar o tratamento de esgoto, o reator anaeróbio do tipo UASB é largamente utilizado no Brasil, porém esta tecnologia apresenta problemas em atender aos limites estabelecidos pela legislação brasileira para lançamento de efluente em termos de nitrogênio amoniacal, fósforo e matéria orgânica, necessitando de pós-tratamento. Diante do potencial para remoção de matéria orgânica e sólidos suspensos pela flotação e da eficiência que a zeólita apresenta na remoção de amônia, a floto-filtração-adsortiva em zeólita pode ser uma promissora alternativa para tratamento deste efluente. Sendo assim, o presente trabalho visou avaliar a eficiência de um sistema de flotação associado à filtração-adsortiva, utilizando zeólita como meio filtrante, para pós-tratamento de efluente de reator UASB. Para tal, foram realizados ensaios de coagulação/floculação, flotação e filtração-adsortiva com um floteste adaptado com filtros adsorventes, utilizando efluente de reator anaeróbio do tipo UASB. Os resultados obtidos neste estudo atenderam aos padrões estabelecidos pela legislação brasileira para lançamento de efluentes em corpos d’água, alcançando 100 % de remoção de nitrogênio amoniacal, permitindo que esse efluente possa ser aproveitado para reuso. Desse modo, o sistema de flotação com a filtração-adsortiva utilizando zeólita como meio filtrante mostrou-se eficiente na remoção de todos os poluentes avaliados. Esta pesquisa teve como principal contribuição a avaliação de alternativa promissora para pós-tratamento de efluente de reator UASB buscando atender as exigências da legislação brasileira e melhorar da qualidade dos efluentes sanitários. ABSTRACT: Among the existing solutions to sewage treatment, the UASB reactor is widely used in Brazil, but this technology has problem to attend the limits established by Brazilian law for discharging of effluent

  17. Retention and failure morphology of prefabricated posts

    DEFF Research Database (Denmark)

    Sahafi, Alireza; Peutzfeldt, Anne; Asmussen, Erik

    2004-01-01

    PURPOSE: This study evaluated the effect of cement, post material, surface treatment, and shape (1) on the retention of posts luted in the root canals of extracted human teeth and (2) on the failure morphology. MATERIALS AND METHODS: Posts of titanium alloy (ParaPost XH), glass fiber (Para...... at 37 degrees C for 7 days, retention was determined by extraction of the posts. Failure morphology of extracted posts was analyzed and quantified stereomicroscopically. RESULTS: Type of luting cement, post material, and shape of post influenced the retention and failure morphology of the posts. Because...

  18. Diseño e Implementación de un Multiprocessor Systems-on-Chip (MPSoC Interconectado por una Networks-on-Chip (NoC

    Directory of Open Access Journals (Sweden)

    Wilson Mauricio Chicaiza

    2013-11-01

    Full Text Available En el presente documento se presenta una breve caracterización de los medios de comunicación empleados en arquitecturas multiprocesadas. Esta caracterización tiene como objetivo principal el mostrar un nuevo modelo de comunicación basado en conmutación de paquetes a los cuales se les denomina como Networks-On-Chip (NoC. Esta publicación muestra una arquitectura de red llamada NoC Hermes, la cual fue interconectada a un Multiprocessor-Systems-on-Chip (MPSoC compuesto de cuatro procesadores MicroBlaze. Está conexión se la realizó gracias al diseño y desarrollo de una Interfaz de Red generada en código VHDL. Por medio de la Interfaz de Red se consiguió que los procesadores MicroBlaze interactúen con los Switches de Hermes a fin de crear una arquitectura multiprocesada interconectada por una NoC. Con el motivo de realizar comparaciones también se creó otra arquitectura de multiprocesadores interconectados por buses. Para ambas arquitecturas se desarrolló una aplicación de Esteganografía enla que existe multiprocesamiento de dos procesadores trabajando simultáneamente. Lamentablemente sobre dicha aplicación no fue posible medir directamente la latencia y el consumo de energía, razón por la cual se utilizó simuladores que permitieron estimar dichas mediciones.

  19. Innovación en tecnologías digitales

    OpenAIRE

    Yepes, Lilit

    2014-01-01

    Los videojuegos ya no deberían ser vistos como el dominio exclusivo de adolescentes, sino más bien como parte de una corriente comercial, de hecho su impacto en la sociedad es de gran alcance. Por ejemplo, las innovaciones líderes en el área de diseño de procesadores, gráficos y la inteligencia artificial están siendo impulsados por la industria de los videojuegos, de esta manera es posible que en la actualidad el software del juego se utilice para entrenar a los soldados para la batalla, y e...

  20. El Word de Microsoft su importancia y mejor utilización

    Directory of Open Access Journals (Sweden)

    Francisco Ficarra

    2015-01-01

    Full Text Available Considera que el editor de textos Word 2000 de Microsoft, es una herramienta indispensable para el trabajo de todo comunicador del siglo XXI. Explica los principales pormenores de la interfaz del Word 2000, cómo beneficiarse del comando AYUDA para resolver las dificultades, cómo dar uniformidad a un documento creado a partir de otros y cómo se crea un estilo propio de plantilla de documento. Da cuenta de la evolución del procesador de textos y de manera didáctica va mostrando la interfaz del Word 2000.

  1. Diseño de un reactor de transesterificación para la obtención de biodiesel a partir de aceites vegetales

    OpenAIRE

    MARSET GIMENO, DAVID

    2016-01-01

    [ES] En este proyecto se pretende que el alumno realice el diseño, montaje y puesta a punto un reactor de transesterificación de laboratorio para la obtención de biodiesel a partir de aceites vegetales, utilizando catálisis básica homogénea. Paralelamente se definirán las técnicas analíticas a emplear para el control de la calidad de los aceites de partida y el seguimiento de los productos de reacción. A partir de los resultados experimentales se realizará el diseño y estimación económica...

  2. Advanced post-irradiation examination techniques for water reactor fuel. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2002-03-01

    The purpose of the meeting was to provide and overview of the status of post-irradiation examination (PIE) techniques for water cooled reactor fuel assemblies and their components with emphasis given to advanced PIE techniques applied to high burnup fuel. Papers presented at the meeting described progress obtained in non-destructive (e.g. dimensional measurements, oxide layer thickness measurements, gamma scanning and tomography, neutron and X-ray radiography, etc.) and destructive PIE techniques (e.g. microstructural studies, elemental and isotopic analysis, measurement of physical and mechanical properties, etc.) used for investigation of water reactor fuel. Recent practice in high burnup fuel investigation revealed the importance of advanced PIE techniques, such as 3-D tomography, secondary ion mass spectrometry, laser flash, high resolution transmission and scanning electron microscopy, image analysis in microstructural studies, for understanding mechanisms of fuel behaviour under irradiation. Importance and needs for in-pile irradiation of samples and rodlets in instrumented rigs were also discussed. This TECDOC contains 20 individual papers presented at the meeting; each of the papers has been indexed separately

  3. Research reactors: a tool for science and medicine; Reactores de investigacion: herramientas para la ciencia y la medicina

    Energy Technology Data Exchange (ETDEWEB)

    Ordonez, Juan [Investigacion Aplicada SE (INVAP), San Carlos de Bariloche (Argentina)

    2001-07-01

    The types and uses of research reactors are reviewed. After an analysis of the world situation, the demand of new research reactors of about 20 MW is foreseen. The experience and competitiveness of INVAP S.E. as designer and constructor of research reactors is outlined and the general specifications of the reactors designed by INVAP for Egypt and Australia are given.

  4. Hydrodynamically induced dryout and post dryout important to heavy water reactors: A yearly progress report

    International Nuclear Information System (INIS)

    Ishii, M.; Revankar, S.T.; Babelli, I.; Lele, S.

    1992-06-01

    Recently, the safety of low pressure liquid cooled nuclear reactors has become a very important issue with reference to the operation of the heavy water reactors at Savannah River Plant. Under accident conditions such as loss-of-flow or loss-of-coolant, these reactors typically encounter unstable two-phase flow which may lead to the occurrence of dryout and subsequent fuel failure. An analytical study using the one-dimensional drift flux model was carried out to investigate the two-phase flow instability for Westinghouse Savannah River Site reactor. The analysis indicates that the first and higher order instabilities exist in the possible transient operational conditions. The instabilities are encountered at higher heat fluxes or lower flow rates. The subcooling has a stabilizing effect except at very low subcooling. An experimental loop has been designed and constructed to study the CBF induced by various flow instabilities. Details of this test loop are presented. Initial test results have been presented. The two-phase flow regimes and hydrodynamic behaviors in the post dryout region have been studied under propagating rewetting conditions. The effect of subcooling and inlet velocity on flow transition as well as on the quench front propagation was investigated. The test liquid was Freon 113 which was introduced into the bottom of the quartz test section whose walls were maintained well above the film boiling temperature of the test liquid, via a transparent heat transfer fluid. The flow regimes observed down stream of the upward moving quench front were the rough wavy, the agitated, and the dispersed droplet/ligaments. A correlation for the flow regime transition between the inverted annular and the dispersed droplet/ligament flow patterns was developed. The correlation showed a marked dependence on the void fraction at the CBF location and hence on the flow regime encountered in the pre-CBF region

  5. Arquitecturas multiprocesador en computación de alto desempeño: software, métricas, modelos y aplicaciones

    OpenAIRE

    De Giusti, Armando Eduardo; Tinetti, Fernando Gustavo; Naiouf, Marcelo; Chichizola, Franco; De Giusti, Laura Cristina; Villagarcía Wanza, Horacio A.; Montezanti, Diego Miguel; Encinas, Diego; Pousa, Adrián; Rodriguez, Ismael Pablo; Rodriguez Eguren, Sebastián; Iglesias, Luciano; Paniego, Juan Manuel; Pi Puig, Martín; Dell'Oso, Matías

    2017-01-01

    Caracterizar las arquitecturas multiprocesador distribuidas enfocadas especialmente a cluster y cloud computing, con énfasis en las que utilizan procesadores de múltiples núcleos (multicores, GPUs y Xeon Phi), con el objetivo de modelizarlas, estudiar su escalabilidad, analizar y predecir performance de aplicaciones paralelas, estudiar el consumo energético y su impacto en la perfomance así como desarrollar esquemas para detección y tolerancia a fallas en las mismas. Profundizar el estudio...

  6. Arquitecturas multiprocesador en HPC: software, métricas y aplicaciones

    OpenAIRE

    De Giusti, Armando Eduardo; Tinetti, Fernando Gustavo; Naiouf, Marcelo; Chichizola, Franco; De Giusti, Laura Cristina; Villagarcía Wanza, Horacio A.; Montezanti, Diego Miguel; Encinas, Diego; Pousa, Adrián; Rodriguez, Ismael Pablo; Eguren, Sebastián; Iglesias, Luciano; Paniego, Juan Manuel; Pi Puig, Martín; Dell'Oso, Matías

    2016-01-01

    Caracterizar las arquitecturas multiprocesador distribuidas enfocadas especialmente a cluster y cloud computing, con énfasis en las que utilizan procesadores de múltiples núcleos (multicores, GPUs y Xeon Phi), con el objetivo de modelizarlas, estudiar su escalabilidad, analizar y predecir performance de aplicaciones paralelas, estudiar el consumo energético y su impacto en la perfomance así como desarrollar esquemas para detección y tolerancia a fallas en las mismas. Profundizar el estudio...

  7. Aplicaciones sobre plataformas de redes neuronales en tiempo real

    OpenAIRE

    Tosini, Marcelo Alejandro; Acosta, Nelson

    2003-01-01

    El presente trabajo pretende el diseño de una metodología para la construcción de aplicaciones basadas en redes neuronales sobre una plataforma Muren. Las aplicaciones se restringen a sistemas de control y reconocimiento de patrones por imágenes. Se describe la arquitectura del sistema de desarrollo Muren, basado en 2 procesadores ZISC de 78 neuronas cada uno, una FPGA Spartan II, bancos de memoria y lógica adicional de comunicación.

  8. The design of electrical heater pins to simulate transient dryout and post-dryout of water reactor fuel

    International Nuclear Information System (INIS)

    Burgess, M.H.; Butcher, A.A.; Sidoli, J.E.A.

    1978-11-01

    A theoretical assessment of indirect and direct filled heater simulations of nuclear reactor fuel pins is described. For reasons of fast temperature response, a direct unfilled heater, with thermocouples buried in the walls, is recommended for studies of Loss-of-Coolant Accidents leading to dryout, post-dryout and rewetting. A design of heater pins, for use in SGHWR or PWR experiments, and compatible with existing 9MW power supplies, is described. Experiments to confirm collapse pressure calculations at 1000 0 C and thermocouple response times are also reported. (author)

  9. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

    The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

  10. Planes de atención de enfermería para pacientes adultos en post-operatorio inmediato de cirugía cardiovascular

    Directory of Open Access Journals (Sweden)

    Parra Vargas Myriam

    1992-06-01

    Full Text Available

    En la Fundación Clínica Shaio de Santafé de Bogotá, se realizó este estudio de tipo descriptivo, utilizando la técnica de seguimiento de casos durante las primeras 24 horas del post-operatorio de cirugía cardíaca sobre ''modelo de planes de atención de enfermería estandarizados y basados en diagnósticos de enfermería prioritarios para el paciente en post-operatorio inmediato de cirugía cardíaca". Para ello se realizó observación directa y valoración seriada (tres en promedio para cada paciente seleccionado a 35 pacientes adultos sometidos a cirugía cardíaca con circulación extracorpórea, atendidos en la unidad de cuidado intensivo quirúrgico durante los cuatro meses dedicados a la recolección de la información y que cumplían con los criterios de selección establecidos. Además de la investigadora principal participaron en esta investigación un grupo de seis (6 estudiantes del programa de especialización en enfermería cardiorespiratorio, quienes se vincularon al estudio como coinvestigadores.

     

  11. vcogmc

    OpenAIRE

    Linares-Barranco, Bernabé; Zamarreño-Ramos, Carlos

    2009-01-01

    El objetivo del proyecto es construir un sistema robótico que incluya sensado procesamiento y actuación motora en forma de impulsos nerviosos tal como se hace en los sistemas biológicos. Para la construcción de este sistema se integrarán retinas y procesadores bioinspirados desarrollados en el IMSE, con algoritmos de actuación motora implementados en FPGA de la Universidad de Sevilla en un robot humanoide ambidiestro de la Universidad Politécnica de Cartagena

  12. Digital convolution

    OpenAIRE

    Camuñas-Mesa, L.; Serrano-Gotarredona, Teresa

    2008-01-01

    El objetivo del proyecto es construir un sistema robótico que incluya sensado procesamiento y actuación motora en forma de impulsos nerviosos tal como se hace en los sistemas biológicos. Para la construcción de este sistema se integrarán retinas y procesadores bioinspirados desarrollados en el IMSE, con algoritmos de actuación motora implementados en FPGA de la Universidad de Sevilla en un robot humanoide ambidiestro de la Universidad Politécnica de Cartagena

  13. 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,...

  14. The post irradiation examination of a sphere-pac (UPu)C fuel pin irradiated in the BR-2 reactor (MFBS 7 experiment)

    International Nuclear Information System (INIS)

    Smith, L.; Aerne, E.T.; Buergisser, B.; Flueckiger, U.; Hofer, R.; Petrik, F.

    1979-09-01

    A pin fuelled with Swiss made (UPu)C microspheres has been successfully irradiated to a peak burn-up of 6% fima in the Belgian BR2 Reactor. The pin, rated up to 95 kW/m, was intact after irradiation and exhibited a peak strain of just over 0.5%. The results of the post irradiation examination are reported. (Auth.)

  15. Fuels and materials research under the high neutron fluence using a fast reactor Joyo and post-irradiation examination facilities

    International Nuclear Information System (INIS)

    Soga, Tomonori; Ito, Chikara; Aoyama, Takafumi; Suzuki, Soju

    2009-01-01

    The experimental fast reactor Joyo at Oarai Research and Development Center (ORDC) of Japan Atomic Energy Agency (JAEA) is Japan's sodium-cooled fast reactor (FR). In 2003, this reactor's upgrade to the 140MWt MK-III core was completed to increase the irradiation testing capability. The MK-III core provides the fast neutron flux of 4.0x10 15 n/cm 2 s as an irradiation test bed for improving the fuels and material of FR in Japan. Three post-irradiation examination (PIE) facilities named FMF, MMF and AGF related to Joyo are in ORDC. Irradiated subassemblies and core components are carried into the FMF (Fuel Monitoring Facility) and conducted nondestructive examinations. Each subassembly is disassembled to conduct some destructive examinations and to prepare the fuel and material samples for further detailed examinations. Fuel samples are sent to the AGF (Alpha-Gamma Facility), and material samples are sent to the MMF (Materials Monitoring Facility). These overall and elaborate data provided by PIE contribute to investigate the irradiation effect and behavior of fuels and materials. This facility complex is indispensable to promote the R and D of FR in Japan. And, the function and technology of irradiation test and PIE enable to contribute to the R and D of innovative fission or fusion reactor material which will be required to use under the high neutron exposure. (author)

  16. Development of PARA-ID Code to Simulate Inelastic Constitutive Equations and Their Parameter Identifications for the Next Generation Reactor Designs

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Lee, J. H.

    2006-03-01

    The establishment of the inelastic analysis technology is essential issue for a development of the next generation reactors subjected to elevated temperature operations. In this report, the peer investigation of constitutive equations in points of a ratcheting and creep-fatigue analysis is carried out and the methods extracting the constitutive parameters from experimental data are established. To perform simulations for each constitutive model, the PARA-ID (PARAmeter-IDentification) computer program is developed. By using this code, various simulations related with the parameter identification of the constitutive models are carried out

  17. Coagulant recovery from water treatment plant sludge and reuse in post-treatment of UASB reactor effluent treating municipal wastewater.

    Science.gov (United States)

    Nair, Abhilash T; Ahammed, M Mansoor

    2014-09-01

    In the present study, feasibility of recovering the coagulant from water treatment plant sludge with sulphuric acid and reusing it in post-treatment of upflow anaerobic sludge blanket (UASB) reactor effluent treating municipal wastewater were studied. The optimum conditions for coagulant recovery from water treatment plant sludge were investigated using response surface methodology (RSM). Sludge obtained from plants that use polyaluminium chloride (PACl) and alum coagulant was utilised for the study. Effect of three variables, pH, solid content and mixing time was studied using a Box-Behnken statistical experimental design. RSM model was developed based on the experimental aluminium recovery, and the response plots were developed. Results of the study showed significant effects of all the three variables and their interactions in the recovery process. The optimum aluminium recovery of 73.26 and 62.73 % from PACl sludge and alum sludge, respectively, was obtained at pH of 2.0, solid content of 0.5 % and mixing time of 30 min. The recovered coagulant solution had elevated concentrations of certain metals and chemical oxygen demand (COD) which raised concern about its reuse potential in water treatment. Hence, the coagulant recovered from PACl sludge was reused as coagulant for post-treatment of UASB reactor effluent treating municipal wastewater. The recovered coagulant gave 71 % COD, 80 % turbidity, 89 % phosphate, 77 % suspended solids and 99.5 % total coliform removal at 25 mg Al/L. Fresh PACl also gave similar performance but at higher dose of 40 mg Al/L. The results suggest that coagulant can be recovered from water treatment plant sludge and can be used to treat UASB reactor effluent treating municipal wastewater which can reduce the consumption of fresh coagulant in wastewater treatment.

  18. Post shut-down decay heat removal from nuclear reactor core by natural convection loops in sodium pool

    Energy Technology Data Exchange (ETDEWEB)

    Rajamani, A. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Sundararajan, T., E-mail: tsundar@iitm.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Prasad, B.V.S.S.S. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Parthasarathy, U.; Velusamy, K. [Nuclear Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2016-05-15

    Highlights: • Transient simulations are performed for a worst case scenario of station black-out. • Inter-wrapper flow between various sub-assemblies reduces peak core temperature. • Various natural convection paths limits fuel clad temperatures below critical level. - Abstract: The 500 MWe Indian pool type Prototype Fast Breeder Reactor (PFBR) has a passive core cooling system, known as the Safety Grade Decay Heat Removal System (SGDHRS) which aids to remove decay heat after shut down phase. Immediately after reactor shut down the fission products in the core continue to generate heat due to beta decay which exponentially decreases with time. In the event of a complete station blackout, the coolant pump system may not be available and the safety grade decay heat removal system transports the decay heat from the core and dissipates it safely to the atmosphere. Apart from SGDHRS, various natural convection loops in the sodium pool carry the heat away from the core and deposit it temporarily in the sodium pool. The buoyancy driven flow through the small inter-wrapper gaps (known as inter-wrapper flow) between fuel subassemblies plays an important role in carrying the decay heat from the sub-assemblies to the hot sodium pool, immediately after reactor shut down. This paper presents the transient prediction of flow and temperature evolution in the reactor subassemblies and the sodium pool, coupled with the safety grade decay heat removal system. It is shown that with a properly sized decay heat exchanger based on liquid sodium and air chimney stacks, the post shutdown decay heat can be safely dissipated to atmospheric air passively.

  19. Development of technologies for nuclear reactors of small and medium sized; Desarrollo de Tecnologias para Reactores Nucleares de pequeno y medio tamano

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-08-15

    This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

  20. Importación de reproductores y nauplios de litopenaeus vannamei para su crianza y exportación al peru, como post-larva

    OpenAIRE

    Villon Noboa, Bolivar; Peñafiel Mena, Ramon; Freire Patiño, Jaime

    2009-01-01

    Existen negocios en acuicultura que pueden ser mucho más rentables de lo que ahora son, sólo con algunas modificaciones que los convertirían en negocios muy atractivos tanto para la inversión nacional como para la extranjera. Tal es el caso de los laboratorios de post-larva de camarón o los laboratorios de maduración, los mismos que, en su mayoría, se encuentran paralizados a lo largo de la costa ecuatoriana. Entre las modificaciones a que se hace referencia destacan, entre otras, la con...

  1. Infraestructura y administración de cómputo paralelo y desarrollo de aplicaciones

    OpenAIRE

    Omar Hernandez Duany; Marlis Fulgueira Camilo; Venus Henry Fuenteseca; Eulises Muñoz Rojas; William Reyes Burunate; Ernesto Insua Suarez

    2015-01-01

    La infraestructura de computación paralela híbrida entre procesadores y tarjetas gráficas es un entorno que permite la ejecución de soluciones paralelas que demandan elevados requisitos de cómputo o que realizan el procesamiento de grandes flujos de datos en tiempo real. En este entorno se han configurado y administrado infraestructuras paralelas para la solución de disímiles problemas que aprovechan las potencialidades que ofrece un clúster de alto rendimiento, construido empleando component...

  2. La transición de un modelo individualista e internista a uno más social y colaborativo en la formación universitaria

    Directory of Open Access Journals (Sweden)

    Luis Ángel Piedra García

    2010-10-01

    Full Text Available En el siguiente documento abordaremos el tema de los procesos de transición de una formación universitaria abundante en argumentos internistas, individualistas y competitivos por uno más social y colaborativo que nos aproxima al modo natural de enseñar y aprender en nuestra especie. Para ello hacemos un análisis general de algunos elementos claves en la formación universitaria analógica que rechaza la idea de que somos procesadores de información.

  3. Desarrollo de un sistema de detección de adelantamiento

    OpenAIRE

    Gómez Gómez, Víctor Manuel; Agüero, Carlos; Matellán Olivera, Vicente; Cañas, José María

    2005-01-01

    El objetivo del Sistema Inteligente de Adelantamiento (SIA) que presentamos en este artículo es la detección automática de situaciones de adelantamiento entre vehículos pesados. Con este proyecto se persigue agilizar esta maniobra aumentando la fluidez del tráfico y mejorando la seguridad. El sistema utiliza una serie de sensores distribuidos por el vehículo y un procesador que está programado para identificar el inicio y fin del adelantamiento, elevando o descendiendo respectivamente el umbr...

  4. Some post operational adjustments to the prototype fast reactor at Dounreay

    International Nuclear Information System (INIS)

    Lunt, A.R.W.

    1979-01-01

    Prior to and during the initial operation of the Prototype Fast Reactor at Dounreay certain features have been considered to be in need of adjustment to provide better operating characteristics. This article describes the work done to support the consequential changes of operational techniques and plant design in the following areas: maintenance of dry conditions at the superheater steam inlets, the temperature control of the reactor roof, and the introduction of a system enabling the reactor to continue running after a turbine trip. (author)

  5. Dissolved methane oxidation and competition for oxygen in down-flow hanging sponge reactor for post-treatment of anaerobic wastewater treatment

    OpenAIRE

    Hatamoto, Masashi; Miyauchi, Tomo; Kindaichi, Tomonori; Ozaki, Noriatsu; Ohashi, Akiyoshi

    2011-01-01

    Post-treatment of anaerobic wastewater was undertaken to biologically oxidize dissolved methane, with the aim of preventing methane emission. The performance of dissolved methane oxidation and competition for oxygen among methane, ammonium, organic matter, and sulfide oxidizing bacteria were investigated using a lab-scale closed-type down-flow hanging sponge (OHS) reactor. Under the oxygen abundant condition of a hydraulic retention time of 2 h and volumetric air supply rate of 12.95 m(3)-air...

  6. Low-enriched research reactor fuel: Post-Irradiation Examinations at SCK-CEN

    International Nuclear Information System (INIS)

    Van den Berghe, S.; Leenaers, A.

    2007-01-01

    Generally, research and test reactors are fuelled with fuel plates instead of pins. In most cases in the past, these plates consisted of high enriched (higher than 95 percent 235 U) UAl 3 powder mixed with a pure Al matrix (called the meat) in between two aluminium alloy plates (the cladding). These plates are then assembled in fuel elements of different designs to fit the needs of the various reactors. Since the 1970's, efforts have been going on to replace the high-enriched, low-density UAl 3 fuel with high-density, low enriched ( 235 U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched materials because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative and the Reduced Enrichment for Research and Test Reactors program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has been obtained with U 3 Si 2 fuel, which is currently used in many research reactors in the world. However, efforts to search for a better replacement have continued and are currently directed towards the U-Mo alloy fuel (7-10 weight percent Mo)

  7. Procesadores de Información : Una Tecnología blanda para el docente

    OpenAIRE

    Orantes, Alfonso

    2013-01-01

    The application of some information processing aids developed within Instructional Psychology are presented. These resources which enhance leaming are very familiar to the teacher. The aim of this aids is to promote an active student participation. To the teacher they represent a "soft technology", easy to use, to prepare and understand, which in developing countries have many advantages over conventional technologies based on expensive harware. As a context, factors afecting text leaming (Le...

  8. Modelo estadístico para la simulación de reactores de lixiviación ácida

    Directory of Open Access Journals (Sweden)

    Mónica Hernández-Rodríguez

    2015-05-01

    Full Text Available Se desarrolló un modelo estadístico que permite simular el comportamiento de la batería de reactores en el proceso de lixiviación ácida y determinar a partir de parámetros operacionales la eficiencia de extracción de níquel y de cobalto. Al realizar las pruebas de validación se obtuvo que más del 95% de los valores determinados por el modelo están dentro de los límites de confianza estimados, sin embargo se observa una tendencia a que el valor calculado se encuentre por debajo del reportado, lo cual se cumple para el 65,79 % y el 61,84 %, de los datos, con relación a la eficiencia de extracción de níquel y cobalto, respectivamente. Se realizó un análisis de sensibilidad paramétrica para establecer la influencia de las variables de operación en el sistema. Se concluye que la sensibilidad depende del nivel de operación del sistema y que las variables más significativas en todos los niveles son: la concentración de magnesio y la de níquel así como la relación ácido - mineral

  9. The utility of different reactor types for the research

    International Nuclear Information System (INIS)

    Stiennon, G.

    1983-01-01

    The report presents a general view of the use of the different belgian research reactor i.e. venus reactor, BR-1 reactor, BR-2 reactor and BR-3 reactor. Particular attention is given to the programmes which is in the interest of international collaboration. In order to reach an efficient utilization of such reactors they require a specialized personnel groups to deal with the irradiation devices and radioactive materials and post irradiation examinations, creating a complete material testing station. (A.J.)

  10. Management of operational events in research reactor

    International Nuclear Information System (INIS)

    Zhong Heping; Yang Shuchun; Peng Xueming

    2001-01-01

    The author describes the tracing management process post-operational event in a research reactor based on nuclear safety code, under the background of the research reactor in Nuclear Power Institute of China. It presorts the definite measures to the event tracing and it up its management factors

  11. Implementación en VHDL de un Detector de Envolvente para demodulación BFSK

    Directory of Open Access Journals (Sweden)

    Karel Toledo de la Garza

    2013-06-01

    Full Text Available El presente artículo aborda el empleo de un bloque Detector de Envolvente para demodular señales BFSK que pueda ser usado en aplicaciones donde se desconoce el tiempo del símbolo de la fuente. Presenta una estructura interna caracterizada por cuatro filtrosdel tipo FIR, que son inherentemente estables y se implementan siempre por una misma ecuación de diferenciasgenérica. El demodulador se configura en lenguaje VHDL con un número variable de coeficientes no especificado de antemano y está sintetizado como un módulo IP con el que se buscareconfigurabilidad. Para validar el  demodulador,se implementa en un circuito FPGA de Xilinx un procesador Microblaze que se comunicacon una PC mediante el puerto serie y se configura con diversos periféricos, tales como la interfaz de comunicación serieRS-232 y el módulo IP del demodulador BFSK especialmente diseñado al efecto. Para gestionar la operación del sistema se desarrolló en la PC un programa en Matlabcon una aplicación gráfica de usuario que incluye el envío y recepción de las señales moduladas y demoduladas por el circuito FPGA, así como el envío de los valores de los coeficientes empleados por los filtros FIR en una determinada aplicación. La solución final permite la demodulación de señales BFSK a través de la interconexión de Matlab con Microblaze y de este con el módulo IP. Se presenta en detalle el modelo VHDL del demodulador, se discuten los resultados alcanzados teniendo en cuenta el efecto de la cuantificación de los coeficientes y se realiza un análisis temporal y de ocupación del circuito FPGA.

  12. Balanceo de un sistema de cosecha mecanizado utilizando simulación de eventos discretos.

    OpenAIRE

    Pablo Aracena; Darío Aedo; Ricardo Landeros

    2010-01-01

    Se analizó un sistema de cosecha mecanizado, operando en una faena a tala rasa de pino radiata y conformado por un feller buncher, un skidder con garra, un procesador y un trineumático. Un Modelo de Simulación de Eventos Discretos (MSED) fue desarrollado con el propósito de balancear el sistema. El proceso mecanizado de madera fue el limitante del sistema de acuerdo con los resultados del estudio de tiempos, entonces este proceso fue apoyado agregando a 3 operadores de motosierra para alcanza...

  13. Diseño y construcción de una despulpadora de frutas horizontal con una capacidad de producción de 250 Kg/h

    OpenAIRE

    Defaz Pallasco, Edison Marcelo; Tuza Cuzco, Fernando Patricio

    2011-01-01

    En la actualidad el Ecuador es un país productor de gran variedad de fruta, es así que parte de las fincas productoras de frutas únicamente se dedican al cultivo y cosecha, para luego venderlos en los mercados nacionales, cuyo valor final es pequeño comparado con el valor de los productos derivados de la fruta. La no existencia de procedimientos que mejoren las condiciones de producción de los procesadores de pulpa, calidad del producto, e incremento en la capacidad competitiva de las unid...

  14. Análisis de la sensibilidad paramétrica en reactores de lecho fijo

    Directory of Open Access Journals (Sweden)

    Hermes A. Rangel Jara

    1992-05-01

    Full Text Available En la búsqueda de los reactores de lecho fijo que ofrezcan una seguridad y permitanmaximizar la conversión -para una determinada longitud del reactor- se analizan los tres arreglos más comunes (paralelo, contracorriente y temperatura constante, con respecto al medio de enfriamiento. Como casos de aplicación se estudiaron la oxidación parcial de O-xileno para producir anhidrido ftálico como producto único en el primer caso y teniendo en cuenta reacciones paralelas y consecutivas para el segundo caso. El sistema de ecuaciones variacionales originado a partir del sistema de ecuaciones diferenciales del modelo del reactor sirve para solucionar el problema de valores de frontera y adicionalmente la sensibilidad paramétrica de las diferentes variables. Mediante un análisis de la sensibilidad paramétrica y de otras ventajas resultantes el arreglo en paralelo puede considerarse como la alternativa más atractiva.

  15. Evaluación del comportamiento hidrodinámico como herramienta para optimización de reactores anaerobios de crecimiento en medio fijo

    Directory of Open Access Journals (Sweden)

    Andrea Pérez

    2008-01-01

    Full Text Available Las condiciones de flujo no ideal en los reactores afectan su desempeño; las causas comunes son cortos circuitos, zonas muertas y recirculación interna por corrientes cinéticas y/o de densidad. En este estudio se optimizó el diseño de un filtro anaerobio a escala real que trata las aguas residuales del proceso de extracción de almidón de yuca, el cual presentaba problemas de represamiento y bajas eficiencias de remoción. La evaluación del comportamiento hidrodinámico inicial mostró la presencia de flujo dual (32% flujo pistón - FP y 37% mezcla completa - CM, zonas muertas (20% y ausencia de cortos circuitos; adicionalmente, la modelación del reactor indicó un grado de dispersión elevado y un comportamiento tendiente a un reactor CM en serie de dos unidades. Con base en estos resultados, se implementaron dos modificaciones en el diseño del reactor: falso fondo y tubería perforada para evacuación de biogás, las cuales permitieron incrementar la fracción de FP (44%, reducir la fracción de zonas muertas (15%, disminuir el Índice de Dispersión (ID e incrementar la tendencia del reactor a un CM en serie de tres unidades, lo que aumentó el tiempo de retención hidráulico (TRH real de 9,6 a 10,2 horas (TRH teórico 12 horas y las eficiencias teóricas de remoción de 73 a 78%.

  16. Nuclear instrumentation for research reactors; Instrumentacion nuclear para reactores nucleares de investigacion

    Energy Technology Data Exchange (ETDEWEB)

    Hofer, Carlos G.; Pita, Antonio; Verrastro, Claudio A.; Maino, Eduardo J. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Unidad de Actividades de Reactores y Centrales Nucleares. Sector Instrumentacion y Control

    1997-10-01

    The nuclear instrumentation for research reactors in Argentina was developed in 70`. A gradual modernization of all the nuclear instrumentation is planned. It includes start-up and power range instrumentation, as well as field monitors, clamp, scram and rod movement control logic. The new instrumentation is linked to a computer network, based on real time operating system for data acquisition, display and logging. This paper describes the modules and whole system aspects. (author). 2 refs.

  17. Simulação numérica aplicada para avaliar o efeito da pré-polimerização no comportamento de reatores tubulares Numerical simulation to evaluate the effect from pre-polymerization on the behavior of tubular reactors

    Directory of Open Access Journals (Sweden)

    André L. Nogueira

    2007-09-01

    Full Text Available O presente estudo utiliza um modelo matemático fenomenológico para simular um sistema de polimerização contínuo em dois estágios. Este sistema é composto por um reator contínuo tipo tanque agitado (CSTR, para pré-polimerização do monômero (primeiro estágio, associado em série a um reator tubular para conduzir a reação até elevados valores de conversão (segundo estágio. Um modelo detalhado, considerando variações axiais e radiais, assim como operação não-isotérmica, foi utilizado para simular o comportamento do reator tubular em diferentes situações. Um modelo de caracterização também foi desenvolvido para fornecer estimativas do peso molecular médio e do índice de polidispersão do polímero. Os resultados mostram que reações de polimerização conduzidas em sistemas contínuos de dois estágios fornecem um polímero com propriedades menos heterogêneas do que um polímero obtido em um sistema reacional composto por apenas um reator tubular. Além disso, quanto maior a viscosidade da mistura reacional alimentada ao reator tubular, mais homogêneo é o polímero obtido.The present study uses a phenomenological model to simulate a continuous, two-stage polymerization process. This system is composed by a continuous stirred tank reactor (CSTR for monomer pre-polymerization (first stage, connected to a tubular reactor (second stage to carry out the reaction up to high conversion values. A comprehensive non-isothermal 2-D model (axial and radial variations was used to predict the tubular reactor behavior. A polymer characterization model was also developed to provide estimates of the polymer average molecular weight and polydispersity. According to the results, polymerization reactions carried out in a continuous two-stage system provide a polymer with less heterogeneous properties than the one obtained in a single tubular reactor. Besides, it is possible to produce a more homogeneous polymer increasing the viscosity

  18. Estudio hidrodinámico de reactores empacados de flujo ascendente(REFA)

    OpenAIRE

    Díaz Marrero, Miguel Ángel; Dueñas Moreno, Jaime; Cabrera Díaz, Ania

    2014-01-01

    En el presente trabajo se emplean las técnicas de estímulo respuesta para estudiar los modelos de flujos de dos reactores tipo REFA, con volúmenes de 3,4 y 6 litros respectivamente, usando tiempos de retención hidráulicos y trazadores diferentes en ambos. Se determinaron las curvas de concentración contra tiempo para ambos reactores y se realizó el análisis comparativo de un grupo de relaciones entre los diferentes tiempos que se obtuvieron en los gráficos. Se aplica con los mismos experiment...

  19. Estudio de criticidad del reactor MSBR con SCALE

    OpenAIRE

    Criado Martín, Alejandro Fernando

    2011-01-01

    El presente proyecto final de carrera se enmarca en el convenio de colaboración entre el Consejo de Seguridad Nuclear (CSN) y la Universitat Politècnica de Catalunya (UPC) para la realización de proyectos en el ámbito de la seguridad nuclear y la protección radiológica. El proyecto estudia la criticidad del reactor Molten Salt Breeder Reactor (MSBR) mediante el código de simulación SCALE. El MSBR es un reactor de sales fundidas concebido y diseñado por ORNL, con una composic...

  20. Dissolved methane oxidation and competition for oxygen in down-flow hanging sponge reactor for post-treatment of anaerobic wastewater treatment.

    Science.gov (United States)

    Hatamoto, Masashi; Miyauchi, Tomo; Kindaichi, Tomonori; Ozaki, Noriatsu; Ohashi, Akiyoshi

    2011-11-01

    Post-treatment of anaerobic wastewater was undertaken to biologically oxidize dissolved methane, with the aim of preventing methane emission. The performance of dissolved methane oxidation and competition for oxygen among methane, ammonium, organic matter, and sulfide oxidizing bacteria were investigated using a lab-scale closed-type down-flow hanging sponge (DHS) reactor. Under the oxygen abundant condition of a hydraulic retention time of 2h and volumetric air supply rate of 12.95m(3)-airm(-3)day(-1), greater than 90% oxidation of dissolved methane, ammonium, sulfide, and organic matter was achieved. With reduction in the air supply rate, ammonium oxidation first ceased, after which methane oxidation deteriorated. Sulfide oxidation was disrupted in the final step, indicating that COD and sulfide oxidation occurred prior to methane oxidation. A microbial community analysis revealed that peculiar methanotrophic communities dominating the Methylocaldum species were formed in the DHS reactor operation. Copyright © 2011 Elsevier Ltd. All rights reserved.

  1. Examples of CEA managements of spent fuels from a prototype power reactor (PHENIX) and from commercial power reactors after post irradiation examinations

    International Nuclear Information System (INIS)

    Guay, P.

    1988-01-01

    CEA gained a good experience in the management of spent fuels from its research or power prototype reactors and of the fuel samples for post irradiation examinations. The solution for these products is the reprocessing. The delay to apply that solution is bound to the disponibility of the reprocessing facilities, and in several cases induce a delayed reprocessing. Only particular and limited fuels are planned to be sent in a definitive storage. The definitive storage is choosen only for a few fuels essentially requiring important modifications of the dissolution process. The treatments and operations on the spent fuels must be carried out following the French safety rules. Long and detailed flowsheet studies are therefore necessary before the setting up of the operations. Generally the cost of the management of limited quantities of fuels, as it is the case here, is high. The flowsheets are established in taking into account, as far as possible, the use of existing facilities, procedures, transport casks

  2. Safety in the Utilization and Modification of Research Reactors. Specific Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-07-15

    This Safety Guide is a revision of Safety Series No. 35-G2 on safety in the utilization and modification of research reactors. It provides recommendations on meeting the requirements for the categorization, safety assessment and approval of research reactor experiments and modification projects. Specific safety considerations in different phases of utilization and modification projects are covered, including the pre-implementation, implementation and post-implementation phases. Guidance is also provided on the operational safety of experiments, including in the handling, dismantling, post-irradiation examination and disposal of experimental devices. Examples of the application of the safety categorization process for experiments and modification projects and of the content of the safety analysis report for an experiment are also provided. Contents: 1. Introduction; 2. Management system for the utilization and modification of a research reactor; 3. Categorization, safety assessment and approval of an experiment or modification; 4. Safety considerations for the design of an experiment or modification; 5. Pre-implementation phase of a modification or utilization project; 6. Implementation phase of a modification or utilization project; 7. Post-implementation phase of a utilization or modification project; 8. Operational safety of experiments at a research reactor; 9. Safety considerations in the handling, dismantling, post-irradiation examination and disposal of experimental devices; 10. Safety aspects of out-of-reactor-core installations; Annex I: Example of a checklist for the categorization of an experiment or modification at a research reactor; Annex II: Example of the content of the safety analysis report for an experiment at a research reactor; Annex III: Examples of reasons for a modification at a research reactor.

  3. Safety in the Utilization and Modification of Research Reactors. Specific Safety Guide

    International Nuclear Information System (INIS)

    2012-01-01

    This Safety Guide is a revision of Safety Series No. 35-G2 on safety in the utilization and modification of research reactors. It provides recommendations on meeting the requirements for the categorization, safety assessment and approval of research reactor experiments and modification projects. Specific safety considerations in different phases of utilization and modification projects are covered, including the pre-implementation, implementation and post-implementation phases. Guidance is also provided on the operational safety of experiments, including in the handling, dismantling, post-irradiation examination and disposal of experimental devices. Examples of the application of the safety categorization process for experiments and modification projects and of the content of the safety analysis report for an experiment are also provided. Contents: 1. Introduction; 2. Management system for the utilization and modification of a research reactor; 3. Categorization, safety assessment and approval of an experiment or modification; 4. Safety considerations for the design of an experiment or modification; 5. Pre-implementation phase of a modification or utilization project; 6. Implementation phase of a modification or utilization project; 7. Post-implementation phase of a utilization or modification project; 8. Operational safety of experiments at a research reactor; 9. Safety considerations in the handling, dismantling, post-irradiation examination and disposal of experimental devices; 10. Safety aspects of out-of-reactor-core installations; Annex I: Example of a checklist for the categorization of an experiment or modification at a research reactor; Annex II: Example of the content of the safety analysis report for an experiment at a research reactor; Annex III: Examples of reasons for a modification at a research reactor.

  4. Effect of surface treatment of prefabricated posts on bonding of resin cement

    DEFF Research Database (Denmark)

    Sahafi, Alireza; Peutzfeld, Anne; Asmussen, Erik

    2004-01-01

    This in vitro study evaluated the effect of various surface treatments of prefabricated posts of titanium alloy (ParaPost XH), glass fiber (ParaPost Fiber White) and zirconia (Cerapost) on the bonding of two resin cements: ParaPost Cement and Panavia F by a diametral tensile strength (DTS) test...... the start of mixing the resin cement, the specimen was freed from the mold and stored in water at 37 degrees C for seven days. Following water storage, the specimen was wet-ground to a final length of approximately 3 mm. The DTS of specimens was determined in a Universal Testing Machine. The bonding...

  5. Ion nitriding post-oxidation as an alternative technique to electrolytic chromium; Nitruracion post-oxidacion ionica como tecnica alternativa al cromado electrolitico

    Energy Technology Data Exchange (ETDEWEB)

    Diaz-Guillen, J. C.; Granda-Gutierrez, E.E.; Campa-Castilla, A.; Perez-Aguilar, S.I.; Garza-Gomez, A.; Candelas-Ramirez, J.; Mendez-Mendez, R. [COMIMSA. Corporacion Mexicana de Investigacion en Materiales S.A. de C.V., Saltillo, Coahuila (Mexico)]. E-mail: jcarlos@comimsa.com

    2010-11-15

    The effect of temperature and processing time during post-oxidation on hardness and corrosion resistance of AISI 1045 samples treated through nitriding and post-oxidation in a pulsed plasma discharge is evaluated in this paper. Also, a comparative analysis of the mechanical properties obtained with the dual nitriding - post oxidation process versus those properties of typical hard chrome coatings was performed with an aim to propose an alternative technique to the processes of galvanic coatings. The latter revealed that the process of ion nitriding and post-oxidation provides similar properties in hardness and improves the corrosion resistance compared to the hard chrome case. It is conclude that the technique of ion nitriding and post-oxidation is a non environmental harmful technology with strong potential to replace highly polluting electroplating techniques for application of hard chrome coatings. [Spanish] En el presente trabajo se evalua el efecto del tiempo y la temperatura de post-oxidacion sobre las propiedades de dureza y resistencia a la corrosion de muestras de acero AISI 1045 sometido al proceso de nitruracion post-oxidacion ionica en plasmas pulsados. Asi mismo, con el objetivo de fundamentar la propuesta de utilizacion de la nitruracion post-oxidacion ionica como una tecnica alternativa a los procesos galvanicos para aplicacion de recubrimientos de cromo duro, se realizo un analisis comparativo de propiedades, evidenciando que, mediante el proceso nitruracion postoxidacion ionica, es posible obtener caracteristicas similares en dureza y mejores en resistencia a la corrosion que las tipicas obtenidas para el cromo duro. Los resultados obtenidos permiten postular una tecnica que no dana al medio ambiente, como lo es la nitruracion post-oxidacion ionica, como candidata potencial para sustituir las tecnicas galvanicas altamente contaminantes para aplicacion de cromo duro.

  6. Sterilization of swine wastewater treated by anaerobic reactors using UV photo-reactors

    Directory of Open Access Journals (Sweden)

    Erlon Lopes Pereira

    2014-09-01

    Full Text Available The use of ultraviolet radiation is an established procedure with growing application forthe disinfection of contaminated wastewater. This study aimed to evaluate the efficiency of artificial UV radiation, as a post treatment of liquid from anaerobic reactors treating swine effluent. The UV reactors were employed to sterilize pathogenic microorganisms. To this end, two photo-reactors were constructed using PVC pipe with100 mm diameter and 1060 mmlength, whose ends were sealed with PVC caps. The photo-reactors were designed to act on the liquid surface, as the lamp does not get into contact with the liquid. To increase the efficiency of UV radiation, photo-reactors were coated with aluminum foil. The lamp used in the reactors was germicidal fluorescent, with band wavelength of 230 nm, power of 30 Watts and manufactured by Techlux. In this research, the HRT with the highest removal efficiency was 0.063 days (90.6 minutes, even treating an effluent with veryhigh turbidity due to dissolved solids. It was concluded that the sterilization method using UV has proved to be an effective and appropriate process, among many other procedures.

  7. Software library of meteorological routines for air quality models; Libreria de software de procedimientos meteorologicos para modelos de dispersion de contaminantes

    Energy Technology Data Exchange (ETDEWEB)

    Galindo Garcia, Ivan Francisco

    1999-04-01

    fundamental para llevar a cabo la evaluacion del impacto de fuentes fijas sobre la calidad del aire. Sin embargo, los modelos requieren cierta informacion relativa a la meteorologia de la zona que se desea modelar. Algunos de estos parametros requeridos se pueden medir directamente, pero otros deben ser estimados a partir de los datos medidos. Debido a esto, para realizar un estudio de modelacion de la contaminacion atmosferica, tambien es necesario contar con un conjunto de procedimientos, relaciones y programas de computo que permitan obtener todos los parametros meteorologicos y micrometeorologicos requeridos como datos de entrada por el modelo de dispersion de contaminantes especifico que se desee utilizar. El objetivo de este trabajo es la identificacion, establecimiento e implementacion de los metodos, relaciones y procedimientos que permiten la estimacion de los parametros meteorologicos y micrometeorologicos requeridos por los modelos de dispersion de contaminantes atmosfericos recomendados por la Agencia de Proteccion Ambiental de los Estados Unidos (US-EPA), a partir de diferentes niveles de disponibilidad de informacion meteorologica primaria. Para ello se realizo un estudio sobre los diferentes modelos de dispersion atmosferica, analizando, en particular, los datos meteorologicos que requieren. Asimismo, se llevo a cabo una caracterizacion de las estaciones meteorologicas mexicanas pertenecientes al Servicio Meteorologico Nacional, de donde se obtuvo informacion referente al tipo y calidad de los datos meteorologicos que producen, a fin de establecer, en particular, una metodologia especifica para la estimacion de los datos meteorologicos necesarios para la modelacion de la calidad del aire en Mexico. Los procedimientos de estimacion desarrollados se organizaron en una libreria de software que permite la integracion de un procesador meteorologico apropiado para cada modelo de dispersion (US-EPA) que se desee utilizar. Los procedimientos de estimacion se validaron

  8. Thermal barrier and support for nuclear reactor fuel core

    International Nuclear Information System (INIS)

    Betts, W.S. Jr.; Pickering, J.L.; Black, W.E.

    1987-01-01

    A nuclear reactor is described having a thermal barrier for supporting a fuel column of a nuclear reactor core within a reactor vessel having a fixed rigid metal liner. The fuel column has a refractory post extending downward. The thermal barrier comprises, in combination, a metallic core support having an interior chamber secured to the metal liner; fibrous thermal insulation material covering the metal liner and surrounding the metallic core support; means associated with the metallic core support and resting on the top for locating and supporting the full column post; and a column of ceramic material located within the interior chamber of the metallic core support, the height of the column is less than the height of the metallic core support so that the ceramic column will engage the means for locating and supporting the fuel column post only upon plastic deformation of the metallic core support; the core support comprises a metallic cylinder and the ceramic column comprises coaxially aligned ceramic pads. Each pad has a hole located within the metallic cylinder by means of a ceramic post passing through the holes in the pads

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

    Directory of Open Access Journals (Sweden)

    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.

  10. Estudio de ecotoxicidad y biodegradabilidad de ibuprofeno en un reactor aerobio de lodos activos de mezcla completa

    OpenAIRE

    Zambrano Flores, Johanna Vanessa

    2013-01-01

    Es importante conocer qué efectos de toxicidad aguda y crónica presenta el Ibuprofeno, así como los posibles efectos tóxicos que a largo plazo puedan producirse sobre la biomasa activa presente en las plantas depuradoras de aguas residuales contaminadas con este compuesto. Para ello, se realizó el montaje de un reactor aerobio de fangos activos de mezcla completa. Primero, se alimentó al reactor únicamente con agua residual sintética para el arranque y operación estacionaria del reactor. Desp...

  11. Post-remedial-action survey report for Kinetic Experiment Water Boiler Reactor Facility, Santa Susana Field Laboratories, Rockwell International, Ventura County, California

    International Nuclear Information System (INIS)

    Wynveen, R.A.; Smith, W.H.; Sholeen, C.M.; Flynn, K.F.; Justus, A.L.

    1981-10-01

    Rockwell International's Santa Susana Laboratories in Ventura County, California, have been the site of numerous federally-funded contracted projects involving the use of radioactive materials. Among these was the Kinetics Experiment Water Boiler (KEWB) Reactor which was operated under the auspices of the US Atomic Energy Commission (AEC). The KEWB Reactor was last operated in 1966. The facility was subsequently declared excess and decontamination and decommissioning operations were conducted during the first half of calendar year 1975. The facility was completely dismantled and the site graded to blend with the surrounding terrain. During October 1981, a post-remedial-action (certification) survey of the KEWB site was conducted on the behalf of the US Department of Energy by the Radiological Survey Group (RSG) of the Occupational Health and Safety Division's Health Physics Section (OHS/HP) of Argonne National Laboratory (ANL). The survey confirmed that the site was free from contamination and could be released for unrestricted use

  12. Final report on in-reactor uniaxial tensile deformation of pure iron and Fe-Cr alloy

    International Nuclear Information System (INIS)

    Singh, B.N.; Xiaoxu Huang; Taehtinen, S.; Moilamen, P.; Jacquet, P.; Dekeyser, J.

    2007-11-01

    Traditionally, the effect of irradiation on mechanical properties of metals and alloys is determined using post-irradiation tests carried out on pre-irradiated specimens and in the absence of irradiation environment. The results of these tests may not be representative of deformation behaviour of materials used in the structural components of a fission or fusion reactor where the materials will be exposed concurrently to displacement damage and external and/or internal stresses. In an effort to evaluate and understand the dynamic response of materials under these conditions, we have recently performed a series of uniaxial tensile tests on Fe-Cr and pure iron specimens in the BR-2 reactor at Mol (Belgium). The present report first provides a brief description of the test facilities and the procedure used for performing the in-reactor tests. The results on the mechanical response of materials during these tests are presented in the form of stress-displacement dose and the conventional stress-strain curves. For comparison, the results of post-irradiation tests and tests carried out on unirradiated specimens are also presented. Results of microstructural investigations on the unirradiated and deformed, irradiated and undeformed, post-irradiation deformed and the in-reactor deformed specimens are also described. During the in-reactor tests the specimens of both Fe-Cr alloy and pure iron deform in a homogeneous manner and do not exhibit the phenomenon of yield drop. An increase in the pre-yield dose increases the yield stress but not the level of maximum flow stress during the in-reactor deformation of Fe-Cr alloy. Neither the in-reactor nor the post-irradiation deformed specimens of Fe-Cr alloy and pure iron showed any evidence of cleared channel formation. Both in Fe-Cr and pure iron, the in-reactor deformation leads to accumulation of dislocations in a homogeneous fashion and only to a modest density. No dislocation cells are formed during the in-reactor or post

  13. Post CHF heat transfer and quenching

    International Nuclear Information System (INIS)

    Nelson, R.A.; Condie, K.G.

    1980-01-01

    This paper describes quantitatively new mechanisms in the post-CHF regime which provide understanding and predictive capability for several current two-phase forced convective heat transfer problems. These mechanisms are important in predicting rod temperature turnaround and quenching during the reflood phase of either a hypothetical loss-of-coolant accident (LOCA) or the FLECHT and Semiscale experiments. The mechanisms are also important to the blowdown phase of a LOCA or the recent Loss-of-Fluid Test (LOFT) experiments L2-2 and L2-3, which were 200% cold leg break transients. These LOFT experiments experienced total core quenching in the early part of the blowdown phase at high (1000 psia) pressures. The mechanisms are also important to certain pressurized water reactor (PWR) operational transients where the reactor may operate in the post-CHF regime for short periods of time. Accurate prediction of the post-CHF heat transfer including core quench during these transients is of prime importance to limit maximum cladding temperatures and prevent cladding deformation

  14. Post irradiation examination of RAF/M steels after fast reactor irradiation up to 33 dpa and < 340 C (ARBOR1). RAFM steels. Metallurgical and mechanical characterisation. Final report for TW2-TTMS-001b, D9

    Energy Technology Data Exchange (ETDEWEB)

    Petersen, C. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). EURATOM, Inst. fuer Materialforschung, Programm Kernfusion

    2010-07-01

    In an energy generating fusion reactor structural materials will be exposed to very high dpa-levels of about 100 dpa. Due to this fact and because fast reactor irradiation facilities in Europe are not available anymore, a reactor irradiation at the State Scientific Center of the Russian Federation with its Research Institute of Atomic Reactors (SSC RIAR), Dimitrovgrad, had been performed in the fast reactor BOR 60 with an instrumented test rig. This test rig contained tensile, impact and Low Cycle Fatigue type specimens used at FZK since many years. Samples of actual Reduced Activation Ferritic/Martensitic (RAF/M) -steels (e.g. EUROFER 97) had been irradiated in this reactor at a lower temperature (< 340 C) up to a damage of 33 dpa. This irradiation campaign was called ARBOR 1. Starting in 2003 one half of these irradiated samples were post irradiation examined (PIE) by tensile testing, low cycle fatigue testing and impact testing under the ISTC Partner Contract 2781p in the hot cells of SSC RIAR. In the post irradiation instrumented impact tests a significant increase in the Ductile to Brittle Transition Temperature as an effect of irradiation has been detected. During tensile testing the strength values are increasing and the strain values reduced due to substantial irradiation hardening. The hardening rate is decreasing with increasing damage level, but it does not show saturation. The low cycle fatigue behaviour of all examined RAF/M - steels show at total strain amplitudes below 1 % an increase of number of cycles to failure, due to irradiation hardening. From these post irradiation experiments, like tensile, low cycle fatigue and impact tests, radiation induced design data, e.g. for verification of design codes, can be generated.

  15. An assessment of post-LOCA radiolytic generation of hydrogen in reactor containment of Indian PHWRs

    International Nuclear Information System (INIS)

    Bose, H.; Shah, G.C.; Dutta, S.

    2002-01-01

    Full text: An event-wise assessment has been carried out for the 220 MWe Indian PHWRs of standardized design, to estimate the post-LOCA release of radiolytic hydrogen inside reactor containment, in absence of steam-zirconium reaction. The assessment is based on (i) the dissolved hydrogen concentration build-up in water corresponding to the decaying gamma dose profile and (ii) the rate of concentration dependent mass-transfer of hydrogen from water to gas-space. It is observed that the total radiolytic hydrogen released is about three times less than that obtained by the conventional method of calculation which assumes the radiolytic yield of hydrogen to be equal to the primary yield G(H 2 ) = 0.44 molecules per 100 eV. It is also seen that a major part (∼90 %) of the total release is due to the spillage of fission product irradiated suppression pool water flowing through the core, followed by moderator and suppression pool surface releases respectively

  16. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  17. Thermal Reactor Safety

    International Nuclear Information System (INIS)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods

  18. Inflación en la Argentina post-convertibilidad: Algunas claves para su explicación

    Directory of Open Access Journals (Sweden)

    Gastón Ángel Varesi

    2010-01-01

    Full Text Available La presente colaboración analiza el proceso inflacionario en el marco de la configuración del nuevo modelo de post-convertibilidad después del agotamiento y crisis del proceso de dolarización. A partir de la explicación que Diamand desarrolló para explicar la dinámica del modelo de substitución de importaciones. Se llega a concluir, que el periodo 2002-2007, después de la devaluación obligada, se ha profundizado la explotación de recursos naturales, hasta cierto punto de escasa racionalidad, acompañado de un descenso del salario real, y el incremento de la productividad del trabajo en beneficio de los agentes concentrados en la cúpula empresaria que opera en Argentina, en gran medida altamente beneficiada por el incremento de los precios internacionales.

  19. Experimental and inspection facilities in post-irradiation of spent fuel pools for the analysis of the behaviour of nuclear fuels in power reactors

    International Nuclear Information System (INIS)

    Ruggirello, G.; Zawerucha, A.

    1992-01-01

    Since the beginning of the Atomic Nuclear Reactors (PHWR) Atucha I and Embalse in Argentine are employed different techniques for the knowing of the fuel bundles performances. It is detailed the facilities on post-irradiation examination. The techniques described are: online measurements, visual inspections, identifications of defective fuels and rods assemblies in spent fuel pools. This controls have made possible the feed-back to the manufactory process and the changes in the manufactory quality controls. (author)

  20. The 'SILOE' reactor at Grenoble, France and associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the SILOE reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  1. The 'OSIRIS' reactor at Saclay, France and available hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the OSIRIS reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  2. Research reactors for power reactor fuel and materials testing - Studsvik's experience

    International Nuclear Information System (INIS)

    Grounes, M.

    1998-01-01

    Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

  3. Status report about the works for the start up of the RA-0 `zero power` nuclear reactor at the Cordoba National University; Estado actual de avance de las tareas para la nueva puesta en marcha del reactor nuclear RA-0 en la Universidad Nacional de Cordoba

    Energy Technology Data Exchange (ETDEWEB)

    Martin, H R; Carballido, C; Oliveras, T

    1992-12-31

    After two years of works at the Cordoba National University for the new start-up of the RA-0 `zero power` nuclear reactor, the results obtained are herein presented. Starting with practically null infrastructure at the beginning, specially in human resources and instrumentation of the reactor, the objectives can be considered satisfactory. The training in work of the future operational staff, the design and the construction of the instrumentation and the fitting of the installations are the principal items described in this paper. An special attention is devoted to the insertion of this type of installation in the university organization, usually not prepared for the quality and control activities, which is necessarily considered in these type of works. (Author). [Espanol] Luego de aproximadamente dos anos de trabajo para la nueva puesta en marcha del REACTOR NUCLEAR RA-0, se han alcanzado los resultados presentados en este trabajo. Partiendo de una infraestructura practicamente inexistente en cuanto a recursos humanos y estado de las instalaciones, los avances logrados son significativos. Comenzando por la capacitacion y el entrenamiento del futuro personal de operacion y pasando por la adecuacion de los equipos y componentes, hasta la confeccion de la documentacion mandatoria, se muestran los aspectos mas destacables de los trabajos realizados. Una atencion especial se dedica a la insercion de una instalacion de este tipo en el ambito universitario, el cual por sus particulares caracteristicas, ha debido ser tenido en cuenta permanentemente para la futura operacion de las instalaciones. (Autor).

  4. Post mortem investigations of the NPP Greifswald WWER-440 reactor pressure vessels

    International Nuclear Information System (INIS)

    Schuhknecht, Jan; Viehrig, Hans-Werner; Weiss, Frank-Peter; Rindelhardt, Udo

    2008-01-01

    The paper presents first results of the post mortem investigations performed at the reactor pressure vessels (RPV) of the Russian WWER-440 type reactors. Trepans were taken from the core weld SN0.1.4 and base metal of the unit 1 RPV. This RPV was annealed after 15 years of operation and operated for two more years. At first the trepan of the core welding seam was investigated by Master Curve (MC) testing. Specimens from 5 locations through the thickness of the welding seam were tested according to ASTM E1921-05. The reference temperature T 0 was calculated with the measured fracture toughness values, K Jc , at brittle failure of the specimen. Generally the K Jc values measured on pre-cracked and side-grooved Charpy size SE(B) specimens of the investigated weld metal follow the course of the Master Curve. The K Jc values show a remarkable scatter. More values than expected lie below the 5% fractile. In addition the MC SINTAP procedure was applied to determine T 0 SINTAP of the brittle fraction of the data set. There are remarkable differences between T 0 and T 0 SINTAP indicating macroscopic inhomogeneous weld metal. The highest T 0 was about 50 C at a distance of 22 mm from the inner surface of the weld. It is 40 K higher compared with T 0 at the inner surface. This is important for the assessment of ductile-to-brittle temperatures measured with sub size Charpy specimens made of weld metal from the inner RPV wall. This material may not represent the most conservative condition. The results presented in this paper show that the Master Curve approach as adopted in the test standard ASTM E1921-05 is applicable to the investigated WWER-440 multilayer weld metal. The results are of direct importance for an advanced WWER-440 RPV integrity assessment. On the other hand the data pool is broadened for a general introduction of the MC based RPV integrity assessment in the national codes. Furthermore general experiences in the cutting of irradiated RPV steels are collected

  5. Post mortem investigations of the NPP Greifswald WWER-440 reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Schuhknecht, Jan; Viehrig, Hans-Werner; Weiss, Frank-Peter; Rindelhardt, Udo [Forschungszentrum Dresden-Rossendorf e.V., Dresden (Germany). Inst. for Safety Research; Keller, Werner [Studsvik GmbH und Co. KG, Stutensee (Germany)

    2008-07-01

    The paper presents first results of the post mortem investigations performed at the reactor pressure vessels (RPV) of the Russian WWER-440 type reactors. Trepans were taken from the core weld SN0.1.4 and base metal of the unit 1 RPV. This RPV was annealed after 15 years of operation and operated for two more years. At first the trepan of the core welding seam was investigated by Master Curve (MC) testing. Specimens from 5 locations through the thickness of the welding seam were tested according to ASTM E1921-05. The reference temperature T{sub 0} was calculated with the measured fracture toughness values, K{sub Jc}, at brittle failure of the specimen. Generally the K{sub Jc} values measured on pre-cracked and side-grooved Charpy size SE(B) specimens of the investigated weld metal follow the course of the Master Curve. The K{sub Jc} values show a remarkable scatter. More values than expected lie below the 5% fractile. In addition the MC SINTAP procedure was applied to determine T{sub 0}{sup SINTAP} of the brittle fraction of the data set. There are remarkable differences between T{sub 0} and T{sub 0}{sup SINTAP} indicating macroscopic inhomogeneous weld metal. The highest T{sub 0} was about 50 C at a distance of 22 mm from the inner surface of the weld. It is 40 K higher compared with T{sub 0} at the inner surface. This is important for the assessment of ductile-to-brittle temperatures measured with sub size Charpy specimens made of weld metal from the inner RPV wall. This material may not represent the most conservative condition. The results presented in this paper show that the Master Curve approach as adopted in the test standard ASTM E1921-05 is applicable to the investigated WWER-440 multilayer weld metal. The results are of direct importance for an advanced WWER-440 RPV integrity assessment. On the other hand the data pool is broadened for a general introduction of the MC based RPV integrity assessment in the national codes. Furthermore general experiences in

  6. Final report on in-reactor uniaxial tensile deformation of pure iron and Fe-Cr alloy

    DEFF Research Database (Denmark)

    Singh, Bachu Narain; Huang, X.; Tähtinen, S.

    , the in-reactor deformation leads to accumulation of dislocations in a homogeneous fashion and only to a modest density. No dislocation cells are formed during the in-reactor or post-irradiation deformation of Fe-Cr and pure iron. Furthermore, in both cases, the slip systems even in the planes with Schmid...... factor value of almost zero get activated during the in-reactor as well as post-irradiation deformation. The main implications of these results are briefly discussed....

  7. POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

    Directory of Open Access Journals (Sweden)

    Vojtěch Caha

    2016-12-01

    Full Text Available The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature. The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.

  8. Hydriding failure in water reactor fuel elements

    International Nuclear Information System (INIS)

    Sah, D.N.; Ramadasan, E.; Unnikrishnan, K.

    1980-01-01

    Hydriding of the zircaloy cladding has been one of the important causes of failure in water reactor fuel elements. This report reviews the causes, the mechanisms and the methods for prevention of hydriding failure in zircaloy clad water reactor fuel elements. The different types of hydriding of zircaloy cladding have been classified. Various factors influencing zircaloy hydriding from internal and external sources in an operating fuel element have been brought out. The findings of post-irradiation examination of fuel elements from Indian reactors, with respect to clad hydriding and features of hydriding failure are included. (author)

  9. BiomaSoft: sistema informático para el monitoreo y evaluación de la producción de alimentos y energía. Parte I

    Directory of Open Access Journals (Sweden)

    J. R Quevedo

    Full Text Available La producción integrada de alimentos y energía en Cuba exige procesar una diversa y voluminosa información para tomar decisiones locales, sectoriales y nacionales, con el propósito de incidir en políticas públicas, por lo que es necesario el apoyo de sistemas automatizados que faciliten el monitoreo y evaluación (ME de la producción integrada de alimentos y energía en municipios cubanos. El objetivo de esta investigación fue identificar las herramientas de diseño del sistema informático BiomaSoft y contextualizar su entorno de aplicación. La metodología de desarrollo de software fue RUP (Proceso Racional Unificado, del inglés Rational Unified Process, con UML (Lenguaje Unificado de Modelado, del inglés Unified Modeling Language como lenguaje de modelado y PHP (Pre-Procesador de Hipertexto, del inglés Hypertext Pre-Processor como lenguaje de programación. El entorno se conceptualizó mediante un modelo de dominio y se especificaron los requisitos funcionales y no funcionales que se debían cumplir, así como el Diagrama de Casos de Uso del sistema, con la descripción de actores. Para el despliegue de BiomaSoft se concibió una configuración basada en dos tipos de nodos físicos (un servidor web y ordenadores clientes, en los municipios que participan en el proyecto «La biomasa como fuente renovable de energía para el medio rural cubano» (BIOMAS-CUBA. Se concluye que el monitoreo y evaluación de la producción integrada de alimentos y energía en las condiciones cubanas puede ser realizado mediante el sistema automatizado BiomaSoft, y a este propósito contribuye la identificación de las herramientas para su diseño y la contextualización de su entorno de aplicación.

  10. The DIDO-reactor at Harwell, U.K. and ancillary hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the DIDO reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  11. Determination of Imidacloprid and metabolites by liquid chromatography with an electrochemical detector and post column photochemical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rancan, M. [Consiglio per la Ricerca e la Sperimentazione in Agricoltura (CRA), Istituto Nazionale di Apicoltura, Via di Saliceto 80, I-40128 Bologna (Italy)]. E-mail: mrancan@inapicoltura.org; Sabatini, A.G. [Consiglio per la Ricerca e la Sperimentazione in Agricoltura (CRA), Istituto Nazionale di Apicoltura, Via di Saliceto 80, I-40128 Bologna (Italy); Achilli, G. [Euroservice s.r.l., Piazza Maggiolini 3A, I-20015 Parabiago, Milan (Italy); Galletti, G.C. [Dipartimento di Chimica ' G.Ciamician' , University of Bologna, Via F. Selmi 2, I-40126 Bologna (Italy)

    2006-01-05

    A procedure for the determination of Imidacloprid and its main metabolites was set up by means of liquid chromatography with an electrochemical detector and post-column photochemical reactor (LC-h{nu}-ED). Sample clean-up was developed for bees, filter paper and maize leaves. Chromatographic conditions were based on a reversed-phase C-18 column operated by phosphate buffer 50 mM/CH{sub 3}CN (80/20, v/v) at pH 2.9. Detection of Imidacloprid and its metabolites was performed at a potential of 800 mV after photoactivation at 254 nm. Compared to conventional techniques such as gas chromatography/mass spectrometry (GC/MS) or LC coupled to other detectors, the present method allows simultaneous trace-level determination of both Imidacloprid (0.6 ng ml{sup -1}) and its main metabolites (2.4 ng ml{sup -1})

  12. Determination of Imidacloprid and metabolites by liquid chromatography with an electrochemical detector and post column photochemical reactor

    International Nuclear Information System (INIS)

    Rancan, M.; Sabatini, A.G.; Achilli, G.; Galletti, G.C.

    2006-01-01

    A procedure for the determination of Imidacloprid and its main metabolites was set up by means of liquid chromatography with an electrochemical detector and post-column photochemical reactor (LC-hν-ED). Sample clean-up was developed for bees, filter paper and maize leaves. Chromatographic conditions were based on a reversed-phase C-18 column operated by phosphate buffer 50 mM/CH 3 CN (80/20, v/v) at pH 2.9. Detection of Imidacloprid and its metabolites was performed at a potential of 800 mV after photoactivation at 254 nm. Compared to conventional techniques such as gas chromatography/mass spectrometry (GC/MS) or LC coupled to other detectors, the present method allows simultaneous trace-level determination of both Imidacloprid (0.6 ng ml -1 ) and its main metabolites (2.4 ng ml -1 )

  13. DINÁMICA DE UN REACTOR DE BIOPELÍCULA ANAEROBIA TIPO INTERCAMBIADOR DE CALOR (RBAIC

    Directory of Open Access Journals (Sweden)

    Carlos Ramiro Escalera Vásquez

    2005-01-01

    Full Text Available Las características dinámicas de un reactor de biopelícula anaerobio tipo intercambiador de calor (RBAIC, usado para el tratamiento de aguas residuales de melazas, fueron estudiadas experimentalmente. Se realizaron experimentos para estudiar la respuesta del reactor a las sobrecargas orgánicas. También se estudiaron los efectos de los cambios de temperatura de las paredes calientes y las temperaturas ambientales, sobre la eficiencia del reactor, bajo condiciones de estado estacionario. Se demostró que el RBAIC es estable ante la ocurrencia de sobrecarga orgánica. Se concluyó que existe una separación de fases microbianas dentro del reactor, en condiciones normales de operación. Es decir, las bacterias acidogénicas predominan en la masa líquida recirculante y las heteroacetogénicas y metanogénicas lo hacen en la biopelícula adherida a las paredes calientes de transferencia de calor, lo cual implica que los cambios de la temperatura de la pared afectan de mayor manera a la eficiencia de remoción, que los cambios de temperatura del entorno. El RBAIC es una configuración  novedosa, con características energéticas favorables para el tratamiento de aguas residuales de la industria alimenticia.

  14. Post-treatment of anaerobic reactor effluent using coagulation/oxidation followed by double filtration.

    Science.gov (United States)

    Cavallini, Grasiele Soares; de Sousa Vidal, Carlos Magno; de Souza, Jeanette Beber; de Campos, Sandro Xavier

    2016-04-01

    This study evaluates the efficacy of a sanitary sewage treatment system, proposing post-treatment of the effluent generated by the upflow anaerobic sludge blanket UASB reactor, through a Fenton coagulation/oxidation ((ferric chloride (FC) or ferrous sulfate (FS) and peracetic acid (PAA)), followed by a double filtration system, composed of a gravel ascending drainage filter and a sand descending filter. Following the assessment of treatability, the system efficiency was evaluated using physicochemical and microbiological parameters. In all treatments performed in the pilot unit, total suspended solids (TSS) were completely removed, leading to a decrease in turbidity greater than 90% and close to 100% removal of total phosphorous. In the FC and PAA combination, the effluent was oxygenated prior to filtration, enabling a more significant removal of biochemical oxygen demand (BOD), which characterizes aerobic degradation even in a quick sand filter. The treatments carried out in the presence of the PAA oxidizing agent showed a more significant bleaching of the effluent. Concerning the microbiological parameters, the simultaneous use of PAA and FC contributed to the partial inactivation of the assessed microorganisms. A 65% recovery of the effluent was obtained with the proposed treatment system, considering the volume employed in filter backwashing.

  15. Jules Horowitz reactor (RJH): its design

    International Nuclear Information System (INIS)

    Dupuy, J.P.

    2002-01-01

    This article presents the design of the new irradiation facility (Jules Horowitz reactor) that is planned to be built on the Cadarache site of Cea. 2 principles have been followed. The first one is based on a physical separation between the systems and activities related to the reactor and the experiments from one hand and the other systems and means dedicated to the treatment of the experimental devices before and after irradiation on the other hand. This first principle implies to build 2 buildings: the reactor building and the nuclear auxiliaries building. Inside the reactor building activities from the reactor itself are separated from those dedicated to experimentation. In order to maximize the efficiency of such a reactor, an important number of simultaneous experiments is expected, which will generate an endless flux of incoming and out-going experiments and as a consequence an important handling work between the different work posts. The second principle aims at easing any handling work without breaking the rules of confinement. The different storing pools, the water pits that lead to the 5 hot cells and the reactor tank will communicate through a water-filled canal that will link the 2 buildings. (A.C.)

  16. IAEA high temperature gas-cooled reactor activities

    International Nuclear Information System (INIS)

    Kendall, J.M.

    2000-01-01

    The IAEA activities on high temperature gas-cooled reactors are conducted with the review and support of the Member states, primarily through the International Working Group on Gas-Cooled Reactors (IWG-GCR). This paper summarises the results of the IAEA gas-cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (authors)

  17. IAEA high temperature gas cooled reactor activities

    International Nuclear Information System (INIS)

    Kendall, J.M.

    2001-01-01

    IAEA activities on high temperature gas cooled reactors are conducted with the review and support of Member States, primarily through the International Working Group on Gas Cooled Reactors (IWGGCR). This paper summarises the results of the IAEA gas cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products, and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (author)

  18. Economic evaluation of fast reactor fuel cycling

    International Nuclear Information System (INIS)

    Hu Ping; Zhao Fuyu; Yan Zhou; Li Chong

    2012-01-01

    Economic calculation and analysis of two kinds of nuclear fuel cycle are conducted by check off method, based on the nuclear fuel cycling process and model for fast reactor power plant, and comparison is carried out for the economy of fast reactor fuel cycle and PWR once-through fuel cycle. Calculated based on the current price level, the economy of PWR one-through fuel cycle is better than that of the fast reactor fuel cycle. However, in the long term considering the rising of the natural uranium's price and the development of the post treatment technology for nuclear fuels, the cost of the fast reactor fuel cycle is expected to match or lower than that of the PWR once-through fuel cycle. (authors)

  19. Bond strength of resin cement to dentin and to surface-treated posts of titanium alloy, glass fiber, and zirconia

    DEFF Research Database (Denmark)

    Sahafi, Alireza; Peutzfeldt, Anne; Asmussen, Erik

    2003-01-01

    PURPOSE: To determine the effect of surface treatments on bond strength of two resin cements (ParaPost Cement and Panavia F) to posts of titanium alloy (ParaPost XH), glass fiber (ParaPost Fiber White), and zirconia (Cerapost), and to dentin. MATERIALS AND METHODS: After embedding, planar surfaces...... of posts (n = 9 to 14) and human dentin (n = 10) were obtained by grinding. The posts received one of three surface treatments: 1. roughening (sandblasting, hydrofluoric acid etching), 2. application of primer (Alloy Primer, Metalprimer II, silane), or 3. roughening followed by application of primer...

  20. Reactor coolant pump seals: improving their performance

    International Nuclear Information System (INIS)

    Pothier, N.E.; Metcalfe, R.

    1986-06-01

    Large CANDU plants are benefitting from transient-resistant four-year reliable reactor coolant pump seal lifetimes, a direct result of AECL's 20-year comprehensive seal improvement program involving R and D staff, manufacturers, and plant designers and operators. An overview of this program is presented, which covers seal modification design, testing, post-service examination, specialized maintenance and quality control. The relevancy of this technology to Light Water Reactor Coolant Pump Seals is also discussed

  1. Tratamiento de aguas industriales mediante reactor biológico de membranas

    OpenAIRE

    Aznar Jiménez, Antonio

    2008-01-01

    El Laboratorio de Ingeniería para el Tratamiento de Aguas de la Universidad Carlos III de Madrid, de investigación y servicios en el tratamiento de aguas residuales, optimiza el diseño y puesta a punto de reactores biológicos de membranas (MBR), indicados para obtener agua depurada de alta calidad y/o aumentar la capacidad de tratamiento.

  2. Effect of temperature on two-phase anaerobic reactors treating slaughterhouse wastewater

    Directory of Open Access Journals (Sweden)

    Simone Beux

    2007-11-01

    Full Text Available The effectiveness of the anaerobic treatment of effluent from a swine and bovine slaughterhouse was assessed in two sets of two-phase anaerobic digesters, operated with or without temperature control. Set A, consisting of an acidogenic reactor with recirculation and an upflow biological filter as the methanogenic phase, was operated at room temperature, while set B, consisting of an acidogenic reactor without recirculation and an upflow biological filter as the methanogenic phase, was maintained at 32°C. The methanogenic reactors showed COD (Chemical Demand of Oxygen removal above 60% for HRT (Hydraulic Retention Time values of 20, 15, 10, 8, 6, 4, and 2 days. When the HRT value in those reactors was changed to 1 day, the COD percentage removal decreased to 50%. The temperature variations did not have harmful effects on the performance of reactors in set A.Avaliou-se a eficiência do tratamento anaeróbio de efluente de matadouro de suínos e bovinos em dois conjuntos de biodigestores anaeróbios de duas fases, operados com e sem controle de temperatura. O conjunto A, formado por um reator acidogênico com recirculação e um filtro biológico de fluxo ascendente, foi operado a temperatura ambiente e o conjunto B, formado por um reator de fluxo ascendente e um filtro biológico de fluxo ascendente, foi mantido a 32°C. Os reatores metanogênicos apresentaram remoção de DQO acima de 60 % para os TRHs de 20, 15, 10, oito, seis, quatro e dois dias. Quando o TRH destes reatores foi mudado para um dia observou-se uma queda da porcentagem de remoção de DQO para 50 %. As variações de temperatura parecem não ter prejudicado o desempenho dos reatores do conjunto A.

  3. Material test reactor fuel research at the BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

    2012-03-15

    required thermal and hydraulic conditions. The availability of a comprehensive set of post irradiation examination facilities on site complements the versatile BR2 reactor to provide a set of high performance tools for MTR fuel qualification. (author)

  4. Large scale visualization on the Cray XT3 using ParaView.

    Energy Technology Data Exchange (ETDEWEB)

    Rogers, David; Geveci, Berk (Kitware, Inc.); Eschenbert, Kent (Pittsburgh Supercomputing Center); Neundorf, Alexander (Technical University of Kaiserslautern); Marion, Patrick (Kitware, Inc.); Moreland, Kenneth D.; Greenfield, John

    2008-05-01

    Post-processing and visualization are key components to understanding any simulation. Porting ParaView, a scalable visualization tool, to the Cray XT3 allows our analysts to leverage the same supercomputer they use for simulation to perform post-processing. Visualization tools traditionally rely on a variety of rendering, scripting, and networking resources; the challenge of running ParaView on the Lightweight Kernel is to provide and use the visualization and post-processing features in the absence of many OS resources. We have successfully accomplished this at Sandia National Laboratories and the Pittsburgh Supercomputing Center.

  5. Photoelastic stress analysis of different prefabricated post-and-core materials.

    Science.gov (United States)

    Asvanund, Pattapon; Morgano, Steven M

    2011-01-01

    The purpose of this study was to investigate stress developed by a combination of a stainless steel post or a fiber-reinforced resin post with a silver amalgam core or a composite resin core. Two-dimensional photoelastic models were used to simulate root dentin. Posts (ParaPost XT and ParaPost-FiberWhite) were cemented with a luting agent (RelyX Unicem). Silver amalgam cores and composite resin cores were fabricated on the posts. Complete crowns were fabricated and cemented on the cores. Each model was analyzed with 2 force magnitudes and in 2 directions. Fringe orders were recorded and compared using ANOVA (p=0.05) and the Scheffe's test. With vertical force, no stress differences occurred among the 4 groups (p=0.159). With a 30-degree force, there was stress differences among the 4 groups (p<0.001). The combination of a fiber-reinforced post and composite resin core could potentially reduce stresses within the radicular dentin when angled loads are applied.

  6. Post-accident fuel relocation and heat removal in the LMFBR

    International Nuclear Information System (INIS)

    Kazimi, M.S.; Tsai, S.S.; Gasser, R.D.

    1976-08-01

    Assessment of the dynamics of post-accident fuel relocation and heat removal is an important aspect of the evaluation of the consequences of a hypothetical accident in an LMFBR. Such an assessment is of particular importance in the evaluation of the post-accident radiological doses around the reactor site. In the present evaluation particular attention is given to the design features of the Clinch River Breeder Reactor Plant (CRBR). Fuel relocation and heat removal, assuming certain conditions have resulted in core disruption, are discussed. The discussion of events and phenomena involved in the relocation processes is centered around the resulting patterns of heat source distribution. The factors influencing fuel relocation and distribution in the inlet and outlet plena of the reactor vessel are discussed. The current technology of in-vessel heat removal is applied to the design of the CRBR reactor. Both fuel debris cooling limits and overall coolant flow in the reactor under natural convection conditions are explored. Some of the uncertainties in ex-vessel fuel behavior are addressed. In particular, the effect of melting the cavity bed on the rate of growth of a molten fuel pool is investigated

  7. The DR 3 reactor at Risoe, Denmark and its associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the DR 2 reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of seven information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  8. 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.

  9. Desempenho de reator anaeróbio-aeróbio de leito fixo no tratamento de esgoto sanitário Performance of anaerobic-aerobic packed-bed reactor in the treatment of domestic sewage

    Directory of Open Access Journals (Sweden)

    Sérgio Brasil Abreu

    2008-06-01

    Full Text Available Este artigo relata a avaliação do desempenho de um reator anaeróbio-aeróbio, preenchido com espuma de poliuretano, para tratamento de esgoto sanitário. Inicialmente, foram testados diferentes tempos de detenção hidráulica (TDH no reator que operou apenas em condições anaeróbias. Em seguida, foi operado o reator combinado anaeróbio-aeróbio. O melhor resultado para o reator em operação exclusivamente anaeróbia foi para o TDH de 10 horas, no qual se conseguiu reduzir a DQO de 389 ± 70 mg/L para 137 ± 16 mg/L. Para o reator anaeróbio-aeróbio, a DQO foi reduzida de 259 ± 69 mg/L para 93 ± 31 mg/L para TDH de 12 h (6 h no estágio anaeróbio e 6 h no aeróbio. A comparação de todos os resultados obtidos evidenciou a importância do pós-tratamento aeróbio na remoção de parcela de matéria orgânica não removida em tratamento unicamente anaeróbio.This paper reports on the performance evaluation of an upflow anaerobic-aerobic reactor, filled with polyurethane matrices, for domestic sewage treatment. Initially, different hydraulic retention times were assayed with the reactor operating exclusively in anaerobic condition. Afterwards, anaerobic-aerobic combined reactor was operated. The anaerobic operation with HRT of 10 h provided the best organic matter removal with COD reduction from 389 ± 70 mg/L to 137 ± 16 mg/L. Under anaerobic-aerobic condition, the COD dropped from 259 ± 69 mg/L to 93 ± 31 mg/L with HRT of 12 h (6 h in anaerobic and 6 h in aerobic stages. Finally, comparing all the obtained results, it was possible to verify the importance of the aerobic post treatment in the removal of part of the organic matter not removed in an exclusively anaerobic treatment.

  10. Final report. U.S. Department of Energy University Reactor Sharing Program

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, John A

    2003-01-21

    Activities supported at the MIT Nuclear Reactor Laboratory under the U.S. DOE University Reactor Sharing Program are reported for Grant DE FG02-95NE38121 (September 16, 1995 through May 31, 2002). These activities fell under four subcategories: support for research at thesis and post-doctoral levels, support for college-level laboratory exercises, support for reactor tours/lectures on nuclear energy, and support for science fair participants.

  11. Caracterización de un descodificador HEVC ejecutándose en un DSP

    OpenAIRE

    Caño Velasco, Jesús Pablo

    2014-01-01

    HEVC es el nuevo estándar de codificación de vídeo que está siendo desarrollado conjuntamente por las organizaciones ITU-T Video Coding Experts Group (VCEG) e ISO/IEC Moving Picture Experts Group (MPEG). Su objetivo principal es mejorar la compresión de vídeo, en relación a los actuales estándares. Es común hoy en día, debido a su flexibilidad para aplicaciones de bajo consumo, diseñar sistemas de descodificación de vídeo basados en un procesador digital de señal (DSP). En la mayoría de...

  12. The FR 2 reactor at Karlsruhe, F.R. Germany and associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the FR 2 reactor and associated hot cell facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  13. Integral physics data for fast-reactor design; Donnees de physique integrale intervenant dans les etudes de reacteur a neutrons rapides; Integral'nye fizicheskie dannye dlya raschetov reaktorov na bystrykh nejtronakh; Datos fisicos integrales para el diseno de reactores rapidos

    Energy Technology Data Exchange (ETDEWEB)

    Loewenstein, W B; Meneghetti, D [Argonne National Laboratory, Argonne, IL (United States)

    1962-03-15

    examinent ces donnees et decrivent leurs domaines d'application. Ils montrent que dans certaines analyses de spectre et d'etat critique, les resultats experimentaux et analytiques sont limites. Ils font des suggestions sur l'orientation des recherches experimentales et analytiques a venir. Elles combleraient le fosse entre la theorie et l'experience qui existe dans les systemes 'connus'. Ces propositions comprennent egalement des suggestions en vue de 'consolider' la physique de modeles theoriques de grands reacteurs surgenerateurs a neutrons rapides. (author) [Spanish] La preparacion del capitulo dedicado a la fisica de los reactores rapidos, en la segunda edicion de la publicacion 'Reactor Physics Constants' que aparecera en breve, exigio la recopilacion de los datos disponibles sobre experimentos integrales. La eleccion de los datos integrales de fisica de los reactores rapidos que se ha de incluir en esa seccion se baso en los dos criterios siguientes: a) que los datos provengan de sistemas relativamente simples que se presten para un analisis teorico sencillo; y b) que se trate de sistemas complejos que representan prototipos o maquetas que ofrecen interes general para el estudio de los reactores de potencia rapidos. Se fijo el primer criterio con la intencion de registrar los datos integrales de aquellos sistemas que tienen una utilidad mas general en la verificacion de los parametros y los procedimientos de calculo de las secciones eficaces. El segundo criterio se basa en la presentacion de los datos corrientes sobre sistemas reales de reactores de potencia reproductores rapidos. Estos son demasiado complicados para permitir un analisis teorico sencillo. Constituyen una demostracion de la complejidad del reactor real si se compara con la instalacion critica de experimentacio n mas esquematica y mas facil de analizar. Los datos fisicos integrales que intrevienen en el diseno de reactores constituyen el resultado de mediciones efectuadas en conjuntos criticos o

  14. EFICIENCIA DE CONSORCIOS MICROBIANOS PARA TRATAMIENTO DE AGUAS RESIDUALES EN UN SISTEMA DE RECIRCULACIÓN ACUÍCOLA

    Directory of Open Access Journals (Sweden)

    IVÁN ANDRÉS SÁNCHEZ O

    Full Text Available Los filtros biológicos viabilizan el reuso de aguas residuales (AR en sistemas de recirculación acuícola (SRA, su desempeño depende entre otros factores, del tipo de filtro, medio soporte e inóculo utilizado. Se evaluaron las eficiencias de diferentes inóculos para el tratamiento de AR provenientes del cultivo de trucha arcoiris mediante biofiltros de flujo ascendente en un SRA. Se utilizó un tanque para cultivo con control de nivel, un filtro con bolsas de lienzo para retención de sólidos y seis biofiltros de diámetro 3” con arena como medio soporte y tiempo de retención hidráulica (TRH de 11 min. Los inóculos utilizados fueron: R1-control: aguas del SRA; R2-lodos estación piscícola; R3-agua laguna aereada relleno sanitario Antanas (RSA; R4-sedimentos de acuarios; R5-lodos laguna aereada RSA; R6-lodos reactor sulfidogénico RSA. No hubo diferencias estadísticamente significativas entre los reactores para remoción de DQO, fósforo, amoniaco y nitritos, cuyas remociones medias fueron de 45,3; 15,1; 4,7 y 27,2% respectivamente. Hubo diferencias estadísticas entre reactores para remoción de color y nitratos, las mejores eficiencias fueron para: color R6:38,8% y R1:37,3%; para nitratos R5:47,3% y R6:42,8%; demostrándose la influencia de consorcios microbianos en los SRA

  15. 1982 annual status report: reactor safety

    International Nuclear Information System (INIS)

    1982-01-01

    This report presents the projects of the Reactor Safety Program at the JRC: 1) Reliability and risk evolution; 2) LWR loss of coolant accident studies; 3) Primary system integrity; 4) LMFBR core accident initiation and transition phase; and, 5) LMFBR accident post disassembly phase

  16. Fuel assemblies for nuclear reactors

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

    In a nuclear fuel assembly, hollow guide posts protrude into a fuel assembly and fitting grill from a biased spring pad with a plunger that moves with the spring pad plugging one end of each of the guide posts. A plate on the end fitting grill that has a hole for fluid discharge partially plugs the other end of the guide post. Pressurized water coolant that fills the guide post volume acts as a shock absorber and should the reactor core receive a major seismic or other shock, the fuel assembly is compelled to move towards a pad depending from a transversely disposed support grid. The pad bears against the spring pad and the plunger progressively blocks the orifices provided by slots in the guide posts thus gradually absorbing the applied shock. After the orifice has been completely blocked, controlled fluid discharge continues through a hole coil spring cooperating in the attenuation of the shock. (author)

  17. Chemical reactor modeling multiphase reactive flows

    CERN Document Server

    Jakobsen, Hugo A

    2014-01-01

    Chemical Reactor Modeling closes the gap between Chemical Reaction Engineering and Fluid Mechanics.  The second edition consists of two volumes: Volume 1: Fundamentals. Volume 2: Chemical Engineering Applications In volume 1 most of the fundamental theory is presented. A few numerical model simulation application examples are given to elucidate the link between theory and applications. In volume 2 the chemical reactor equipment to be modeled are described. Several engineering models are introduced and discussed. A survey of the frequently used numerical methods, algorithms and schemes is provided. A few practical engineering applications of the modeling tools are presented and discussed. The working principles of several experimental techniques employed in order to get data for model validation are outlined. The monograph is based on lectures regularly taught in the fourth and fifth years graduate courses in transport phenomena and chemical reactor modeling, and in a post graduate course in modern reactor m...

  18. Thermohydraulic relationships for advanced water cooled reactors

    International Nuclear Information System (INIS)

    2001-04-01

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  19. Thermohydraulic relationships for advanced water cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-04-01

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  20. Viabilidade económica da implementação de um reactor nuclear para a produção de energia eléctrica em Portugal

    OpenAIRE

    Pedro, Miguel António de Morais

    2012-01-01

    O presente trabalho tem como objectivo avaliar economicamente e determinar a viabilidade da implementação de um reactor nuclear para produção de energia eléctrica. Faz-se uma abordagem a aspectos da energia nuclear no mundo e em particular a energia nuclear na união europeia, faz-se uma análise sobre a estrutura do sector nuclear em Espanha e o futuro da energia no mundo. É realizada uma análise sobre a energia nuclear em Portugal, são abordados aspectos como o planeamento energético, a local...

  1. A Downflow Hanging Sponge (DHS) reactor for faecal coliform removal from an Upflow Anaerobic Sludge Bed (UASB) effluent

    NARCIS (Netherlands)

    Yaya Beas, R.E.; Kujawa-Roeleveld, K.; Lier, van J.B.; Zeeman, G.

    2015-01-01

    This research was conducted to study the faecal coliforms removal capacity of Downflow Hanging Sponge (DHS) reactors as a post-treatment for an Upflow Anaerobic Sludge Blanket (UASB) reactor. Three long-term continuous lab-scale DHS reactors i.e. a reactor with cube type sponges without

  2. Operação de filtros biológicos percoladores pós-reatores UASB sem a etapa de decantação secundária Operation of trickling filters post-UASB reactors without the secondary sedimentation stage

    Directory of Open Access Journals (Sweden)

    Paulo Gustavo Sertório de Almeida

    2011-09-01

    Full Text Available A pesquisa teve por objetivo avaliar filtros biológicos percoladores (FBP pós-reatores UASB operando sem a etapa de decantação secundária, em termos da remoção da demanda bioquímica e química de oxigênio (DBO e DQO e sólidos suspensos totais (SST. O aparato experimental consistia em um reator UASB que alimentava quatro FBP em paralelo, preenchidos com diferentes materiais suporte. O reator UASB operou em regime hidráulico permanente, e três condições operacionais foram impostas aos FBP durante o período experimental. Em geral, os sistemas UASB/FBP foram capazes de promover o atendimento aos padrões de lançamento. Em condições de baixas cargas orgânicas volumétricas (COV, o uso de materiais de enchimento de maior área superficial específica não proporcionou ganhos expressivos em termos de desempenho. Contudo, o uso de meio suporte baseado em espumas de poliuretano propiciou melhoria significativa na qualidade do efluente final. O uso de sistemas UASB/FBP sem decantadores secundários parece ser uma promissora alternativa para a simplificação operacional da tecnologia, e uma importante estratégia para o tratamento de efluentes domésticos em países em desenvolvimento. No entanto, o sucesso do emprego desta tecnologia fica condicionado ao correto gerenciamento do lodo anaeróbio do reator UASB, a fim de que sejam evitadas sobrecargas nos FBP.The research aimed at evaluating the operation of trickling filters (TF post-UASB reactors without the secondary sedimentation stage, in terms of biochemical and chemical oxygen demand (BOD and COD and total suspended solids (TSS removal. The experimental apparatus consisted of one UASB reactor followed by four TF in parallel, each one filled with a different packing media. The UASB reactor was operated at a permanent hydraulic regime, while three operational conditions were imposed to the TF during the experimental period. In general, the UASB/TF systems were able to comply with

  3. The FRJ 1 reactor (MERLIN) at Juelich, F.R. Germany and associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the FRJ 1 reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  4. SIMULACION DE LA PUESTA EN MARCHA DE UN REACTOR DE BIOPELÍCULA ANAEROBIA TIPO INTERCAMBIADOR DE CALOR

    Directory of Open Access Journals (Sweden)

    Ramiro Escalera Vásquez

    2009-01-01

    Full Text Available Se ha desarrollado un  modelo de reactor que considera la separación de fases microbianas dentro de un reactor anaerobio tipo intercambiador de calor, donde las bacterias acidogénicas predominan en la masa líquida recirculante y las heteroacetogénicas y metanogénicas lo hacen en la biopelícula adherida a las paredes. El modelo considera también las resistencias difusionales a la transferencia de masa ocasionadas por la capa laminar y la biopelícula. También se consideran las reacciones paralelas y consecutivas propias de la degradación anaerobia de compuestos orgánicos fácilmente biodegradables, por ejemplo, residuos industriales de altas concentraciones de carbohidratos. El modelo de reactor y las ecuaciones pseudo-analíticas para la estimación de los factores de efectividad, desarrolladas para otro tipo de bioreactores anaerobios tales como lechos empacados y fluidizados, pueden utilizarse para estimar la eficacia y evaluar el funcionamiento de un Reactor de Biopelícula Anaerobia tipo Intercambiador de Calor (RBAIC . En este trabajo se ha verificado que los resultados del modelo concuerdan con los resultados experimentales de la eficacia y funcionamiento del RBAIC, dentro del periodo de puesta en marcha.

  5. Glucose isomerization in simulated moving bed reactor by Glucose isomerase

    Directory of Open Access Journals (Sweden)

    Eduardo Alberto Borges da Silva

    2006-05-01

    Full Text Available Studies were carried out on the production of high-fructose syrup by Simulated Moving Bed (SMB technology. A mathematical model and numerical methodology were used to predict the behavior and performance of the simulated moving bed reactors and to verify some important aspects for application of this technology in the isomerization process. The developed algorithm used the strategy that considered equivalences between simulated moving bed reactors and true moving bed reactors. The kinetic parameters of the enzymatic reaction were obtained experimentally using discontinuous reactors by the Lineweaver-Burk technique. Mass transfer effects in the reaction conversion using the immobilized enzyme glucose isomerase were investigated. In the SMB reactive system, the operational variable flow rate of feed stream was evaluated to determine its influence on system performance. Results showed that there were some flow rate values at which greater purities could be obtained.Neste trabalho a tecnologia de Leito Móvel Simulado (LMS reativo é aplicada no processo de isomerização da glicose visando à produção de xarope concentrado de frutose. É apresentada a modelagem matemática e uma metodologia numérica para predizer o comportamento e o desempenho de unidades reativas de leito móvel simulado para verificar alguns aspectos importantes para o emprego desta tecnologia no processo de isomerização. O algoritmo desenvolvido utiliza a abordagem que considera as equivalências entre as unidades reativas de leito móvel simulado e leito móvel verdadeiro. Parâmetros cinéticos da reação enzimática são obtidos experimentalmente usando reatores em batelada pela técnica Lineweaver-Burk. Efeitos da transferência de massa na conversão de reação usando a enzima imobilizada glicose isomerase são verificados. No sistema reativo de LMS, a variável operacional vazão da corrente de alimentação é avaliada para conhecer o efeito de sua influência no

  6. The 33 years of research reactors in JAERI

    International Nuclear Information System (INIS)

    1990-11-01

    The development and utilization of atomic energy in Japan began with the installation of JRR-1 reactor which attained the criticality in August, 1957, thus the third fire was lighted for the first time in Japan. JRR-2 was constructed as a full scale versatile research reactor, which attained the criticality in October, 1960, and since 1962, it has accomplished the role of the reactor for joint utilization. JRR-3 is the first reactor made in Japan by concentrating Japanese technologies in it, to develop and improve Japanese atomic energy technology. It attained the criticality in September, 1962, and has been used as a versatile research reactor. In 1960, Research Reactor Management Department was founded. JRR-4 was constructed as the research reactor for shielding for developing a nuclear-powered ship, which attained the criticality in January, 1965. The first hot laboratory in Japan for carrying out the post-irradiation test on the fuel and materials irradiated in these research reactors was installed in 1961. The JRR-1 was stopped in September, 1968, and is used as the commemorative exhibition hall. The JRR-3 was reconstructed, and attained the criticality in March, 1990, using 20 % enriched uranium fuel. The course of the research reactors for 33 years is reported. (K.I.)

  7. Post-accident core coolability of light water reactors

    International Nuclear Information System (INIS)

    Michio, I.; Teruo, I.; Tomio, Y.; Tsutao, H.

    1983-01-01

    A study on post-accident core coolability of LWR is discussed based on the practical fuel failure behavior experienced in NSRR, PBF, PNS and others. The fuel failure behavior at LOCA, RIA and PCM conditions are reviewed, and seven types of fuel failure modes are extracted as the basic failure mechanism at accident conditions. These are: cladding melt or brittle failure, molten UO 2 failure, high temperature cladding burst, low temperature cladding burst, failure due to swelling of molten UO 2 , failure due to cracks of embrittled cladding for irradiated fuel rods, and TMI-2 core failure. The post-accident core coolability at each failure mode is discussed. The fuel failures caused actual flow blockage problems. A characteristic which is common among these types is that the fuel rods are in the conditions violating the present safety criteria for accidents, and UO 2 pellets are in melting or near melting hot conditions when the fuel rods failed

  8. Diseño de un microreactor para la producción de hidrógeno a partir de alcoholes

    OpenAIRE

    Griffon, Fabien

    2006-01-01

    El objetivo de este trabajo ha sido diseñar un reactor químico para producir hidrógeno a partir de etanol. El reactor funciona gracias a dos reacciones catalizadas que tienen lugar paralelamente: • el reformado: CH3-CH=O + 3 H2O → 2 CO2 + 5 H2 • la oxidación: CH3-CH2-OH + 2 O2 → 2 CO2 + 3 H2O La reacción de reformado transforma el alcohol en hidrógeno, y la de oxidación, que es muy exotermica, suministra energía para la reacción de reformado. El reactor es de paredes catalít...

  9. Tritium release from lithium silicate and lithium aluminate, in-reactor and out-of-reactor

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.

    1976-09-01

    Studies were conducted to determine the generation and evolution of tritium and helium in lithium aluminate (LiAlO 2 ) and lithium silicate (Li 2 SiO 3 ) by the reaction: Li 6 + n → 4 He + T. Targets were irradiated 4.4 days in the K-West Reactor snout facility. (Silicate GVR* approximately 2.0 cc/cc; aluminate GVR approximately 1.4 cc/cc.) Gas release in-reactor was determined by post-irradiation drilling experiments on aluminum ampoules containing silicate and aluminate targets. In-reactor tritium release (at approximately 100 0 C) was found to decrease linearly with increasing target density. Tritium released in-reactor was primarily in the noncondensible form (HT and T 2 ), while in laboratory extractions (300-1300 0 C), the tritium appeared primarily in the condensible form (HTO and T 2 O). Concentrations of HT (and presumably HTO) were relatively high, indicating moisture pickup in canning operations or by inleakage of moisture after the capsule was welded. Impurities in extracted gases included H 2 O, CO 2 , CO, O 2 , H 2 , NO, SO 2 , SiF 4 and traces of hydrocarbons

  10. Training nuclear power plant personnel on SR-O reactor

    International Nuclear Information System (INIS)

    Cerny, K.; Boucek, F.; Kveton, M.; Prokopec, Z.; Fleischhans, J.

    1983-01-01

    The SR-O reactor is an experimental pool-type reactor with a maximum output of 1 MW and maximum thermal neutron flux density of 5.3x10 13 m -2 s -1 . The reactor is described in detail and its specifications are given. The protection and control systems of the reactor permit both manual and automatic operation. The reactor is used for training courses for nuclear power plant operators and for post-graduate study courses for other specialists. Intensive courses for 4 to 6 persons take 15 to 20 days. The course is adjusted to the results of introductory theoretical tests. An optimal teaching method has been developed based on the flowchart algorithmic method, dividing activities into operations (manipulations with controls, issuing commands, making records, etc.) and decision making (information reception and processing). (M.D.)

  11. Aplicación de dos nuevos algoritmos para agrupar resultados de búsquedas en sistemas de catálogos públicos en línea (OPAC

    Directory of Open Access Journals (Sweden)

    Andrés Marín

    2008-01-01

    Full Text Available Con la facilidad que da la Internet y, en particular la Web, cada día es más fácil acceder a nuevas fuentesde información puestas a disposición en cualquier lugar del mundo. Los usuarios buscan informaciónespecífica de acuerdo a sus necesidades particulares, a través de la Web. Ellos pueden hacer búsquedasya sea mediante motores de búsqueda tales como Google o Yahoo!, o también mediante bases de datosparticulares de bibliotecas o sistemas de información. Sin embargo, los resultados de consultas enmotores de búsqueda, sistemas de catálogos de acceso público en línea, y en general sistemas deconsulta en la Web, pueden saturar a un usuario por la abundancia de resultados, causando pérdida deefectividad del sistema de búsqueda. Para resolver este problema, la investigación "Agrupamiento deresultados obtenidos de consultas distribuidas en sistemas de catálogos públicos en línea (OPAC",de la que se deriva este artículo, propone dos algoritmos de agrupamiento de resultados orientados asistemas en línea concurrentes, con características de bajo consumo de ciclos de procesador y memoria,los cuales se usan en un prototipo de software.

  12. International benchmark on the natural convection test in Phenix reactor

    International Nuclear Information System (INIS)

    Tenchine, D.; Pialla, D.; Fanning, T.H.; Thomas, J.W.; Chellapandi, P.; Shvetsov, Y.; Maas, L.; Jeong, H.-Y.; Mikityuk, K.; Chenu, A.; Mochizuki, H.; Monti, S.

    2013-01-01

    Highlights: ► Phenix main characteristics, instrumentation and natural convection test are described. ► “Blind” calculations and post-test calculations from all the participants to the benchmark are compared to reactor data. ► Lessons learned from the natural convection test and the associated calculations are discussed. -- Abstract: The French Phenix sodium cooled fast reactor (SFR) started operation in 1973 and was stopped in 2009. Before the reactor was definitively shutdown, several final tests were planned and performed, including a natural convection test in the primary circuit. During this natural convection test, the heat rejection provided by the steam generators was disabled, followed several minutes later by reactor scram and coast-down of the primary pumps. The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) named “control rod withdrawal and sodium natural circulation tests performed during the Phenix end-of-life experiments”. The overall purpose of the CRP was to improve the Member States’ analytical capabilities in the field of SFR safety. An international benchmark on the natural convection test was organized with “blind” calculations in a first step, then “post-test” calculations and sensitivity studies compared with reactor measurements. Eight organizations from seven Member States took part in the benchmark: ANL (USA), CEA (France), IGCAR (India), IPPE (Russian Federation), IRSN (France), KAERI (Korea), PSI (Switzerland) and University of Fukui (Japan). Each organization performed computations and contributed to the analysis and global recommendations. This paper summarizes the findings of the CRP benchmark exercise associated with the Phenix natural convection test, including blind calculations, post-test calculations and comparisons with measured data. General comments and recommendations are pointed out to improve future simulations of natural convection in SFRs

  13. Review of fast reactor activities

    Energy Technology Data Exchange (ETDEWEB)

    Balz, W [Commission of the European Communities, Brussels (Belgium)

    1978-07-01

    The Commission of the European Communities continued its activities on the following lines: activities aimed at preparing for commercialization of fast breeder reactors which are essentially performed in the frame of Fast Reactor Coordinating Committee (FRCC); the execution of its own research program in the Joint Research Center. The report covers activities of the FRCC, of the Safety Working Group (SWG), the Whole Core Accident Code (WAC) subgroup, Containment (CONT) subgroup, Codes and Standards Working Group (CSWG). Research and development activities are concerned with LMFBR safety, subassembly thermal hydraulics, fuel-coolant interactions, post-accident heat removal, dynamic load response, safety related material properties, utilization limits of fast breeder fuels, plutonium and actinide aspects of nuclear fuel cycle.

  14. Review of fast reactor activities

    International Nuclear Information System (INIS)

    Balz, W.

    1978-01-01

    The Commission of the European Communities continued its activities on the following lines: activities aimed at preparing for commercialization of fast breeder reactors which are essentially performed in the frame of Fast Reactor Coordinating Committee (FRCC); the execution of its own research program in the Joint Research Center. The report covers activities of the FRCC, of the Safety Working Group (SWG), the Whole Core Accident Code (WAC) subgroup, Containment (CONT) subgroup, Codes and Standards Working Group (CSWG). Research and development activities are concerned with LMFBR safety, subassembly thermal hydraulics, fuel-coolant interactions, post-accident heat removal, dynamic load response, safety related material properties, utilization limits of fast breeder fuels, plutonium and actinide aspects of nuclear fuel cycle

  15. Reactor de película líquida descendente para la sulfonación de ésteres metílicos con trióxido de azufre

    Directory of Open Access Journals (Sweden)

    Jesús Alfonso Torres Ortega

    2009-09-01

    Full Text Available Se realizó un conjunto de experimentos de sulfonación de dodecilbenceno (DDB y ésteres metílicos (ME derivados de la esteari- na hidrogenada de palma, con SO3 gaseoso desorbido del óleum, en un reactor de sulfonación en película líquida descendente a escala banco de 40 cm de longitud y ½ pulgada de diámetro interno. Mediante titulaciones volumétricas se determinaron los porcentajes de materia sulfonada y contenido de ácido sulfúrico, así como el porcentaje de aceite libre mediante extracciones con éter de petróleo. La funcionalidad del reactor se verificó efectuando ensayos a condiciones reportadas por Gutiérrez y cola- boradores para dodecilbenceno sulfonado (DDBS, para lo cual fueron determinadas las técnicas de análisis en el Laboratorio de Ingeniería Química (LIQ de la Universidad Nacional de Colombia, sede Bogotá, con el acompañamiento de la empresa Química Básica Colombiana (Caloto, Cauca. Finalmente, se procedió a evaluar la influencia de diferentes variables de proce- so sobre la sulfonación de la mezcla de ésteres metílicos. Los resultados obtenidos en el sulfonador se ajustaron por regresión li- neal múltiple a ecuaciones empíricas, obteniendo expresiones que muestran de forma directa el efecto de variables como la re- lación molar SO3/ME, concentración de SO3 en la corriente gaseosa y flujo másico de ME.

  16. MTR fuel plate qualification in OSIRIS reactor

    International Nuclear Information System (INIS)

    Sacristan, P.; Boulcourt, P.; Naury, S.; Marchard, L.; Carcreff, H.; Noirot, J.

    2005-01-01

    Qualification of new MTR fuel needs the irradiation in research reactors under representative neutronic, heat flux and thermohydraulic conditions. The experiments are performed in France in the OSIRIS reactor by irradiating MTR full size fuel plates in the IRIS device located in the reactor core. The fuel plates are easily removed from the device during the shutdown of the reactor for performing thickness measurements along the plates by means of a swelling measurement device. Beside the calculation capabilities, the experimental platform includes: the ISIS neutron mock-up for the measurement of neutron flux distribution along the plates; the γ spectrometry for the purpose of measuring the activities of the radionuclides representative of the power and the burnup and to compare with the neutronic calculation. Owing to the experience feedback, a good agreement is observed between calculation and measurement; destructive post irradiation examinations in the LECA facility (Cadarache). New irradiations with the IRIS device and at higher heat flux are under preparation for qualification of MTR fuels. (author)

  17. Concept of the new generation high safety liquid metal reactor (LMFR)

    International Nuclear Information System (INIS)

    Alekseev, P.N.; Zverkov, Y.A.; Morozov, A.G.; Orlov, V.V.; Ponomarev-Stepnoi, N.N.; Proshkin, A.A.; Slesarev, I.S.; Subbotin, S.A.

    1988-01-01

    The comparative analysis of the inner stability of the liquid metal reactors to severe accidents was made using the asymptotic reactivity balance. The group of the BN-reactors, Superphenix, IFR, LMFR were considered. This paper lists the characteristics of the reactors, used in the self-protectiveness analysis. The authors present the maximum coolant temperatures in post-accident asymptotic state for IFRs as on of the possible designs of a high safety fast reactor with metal fuel, U-Pu-Zr and LMFR. As is known, these values are very important for assessment of the ATWS accidence consequences. The authors consider the following situations and their combinations: loss of reactor coolant flow-LOFWS, loss of heat sink-LOHSWS, uncontrolled reactor sodium overcooling (down to the freezing point)-OVCWS, uncontrolled excess reactivity insertion-TOPWS. The calculation results demonstrate a high stability of the IFR and LMFR reactors to the most severe accidence sequences

  18. Comparative analysis of power conversion cycles optimized for fast reactors of generation IV; Analisis comparativo de ciclos de conversion de potencia optimizados para reactores rapidos de generacion IV

    Energy Technology Data Exchange (ETDEWEB)

    Perez Pichel, G. D.

    2011-07-01

    For the study, which is presented here, has been chosen as the specific parameters of each reactor, which are today the three largest projects within generation IV technology development: ESFR for the reactor's sodium, LEADER for the lead reactor's and finally, GoFastR in the case of reactor gas-cooled.

  19. Diseño del sistema de control de un fermentador para elaboración doméstica de cerveza

    OpenAIRE

    MARTÍNEZ TOMÁS, MIGUEL

    2015-01-01

    [ES] En el presente trabajo se pretende integrar la tecnología de controladores PIC de bajo costo conocidos como Arduino con el proceso tradicional de fabricación de cerveza de fermentación alta, de cara a la fabricación de reactores domésticos de fermentación que simplifiquen la entrada al mundo del homebrewing al público. Para ello se ha diseñado un código en el lenguaje nativo de la plataforma capaz de controlar un reactor piloto fabricado a efecto de pruebas para este proye...

  20. Investigation on cause of outage of Wide Range Monitor (WRM) in High Temperature engineering Test Reactor (HTTR). Post Irradiation Examination (PIE) toward investigation of the cause

    International Nuclear Information System (INIS)

    Shinohara, Masanori; Saito, Kenji; Takada, Shoji; Ishimi, Akihiro; Katsuyama, Kozo; Motegi, Toshihiro

    2012-08-01

    An event, in which one of WRMs were disabled to detect the neutron flux in the reactor core, occurred during the period of reactor shut down of HTTR in March, 2010. The actual life time of WRM was unexpectedly shorter than the past developed life time. Investigation of the cause of the outage of WRM toward the recovery of the life time up to the past developed life is one of the issues to develop the technology basis of High Temperature Gas cooled Reactor (HTGR). Then, two experimental investigations were carried out to reveal the cause of the outage by specifying the damaged part causing the event in the WRM. The one is a post irradiation examination using the X-ray computed tomography scanner in Fuels Monitoring Facility (FMF) to specify the damaged part in the WRM. The other is an experiment using a mock-up simulating the WRM fabricated by the fabricator. The characteristic impedance of the damaged WRM was measured by Time Domain Reflectometry, which was compared with that of the mock-up, which could narrow down the damaged part in the WRM. This report summarized the results of the PIE and the experimental investigation using the mock-up to reveal the cause of outage of WRM. (author)

  1. Graphics and control for in-reactor operations

    International Nuclear Information System (INIS)

    Smith, A.L.

    1996-01-01

    A wide range of manipulator systems has been developed to carry out remotely operated inspection, repair and maintenance tasks at the Magnox reactors in the United Kingdom. A key factor in the improvement of these systems in recent years has been the extensive use of computer graphics as a real-time aid to the manipulator operator. This is exemplified by the reactor pressure vessel inspection work at the Bradwell reactor which is described in detail. The graphics sub-system of the control system for the manipulator plays a unique and wide-ranging role. The 3D modelling and simulation capability of the IGRIP software has contributed to the conceptual design, detailed path planning, rehearsal support, public relations, real-time manipulator display, post inspection documentation and quality assurance. (UK)

  2. Improving plant availability by predicting reactor trips

    International Nuclear Information System (INIS)

    Frank, M.V.; Epstein, S.A.

    1986-01-01

    Management Ahnalysis Company (MAC) has developed and applied two complementary software packages called RiTSE and RAMSES. Together they provide an mini-computer workstation for maintenance and operations personnel to dramatically reduce inadvertent reactor trips. They are intended to be used by those responsible at the plant for authorizing work during operation (such as a clearance coordinator or shift foreman in U.S. plants). They discover and represent all components, processes, and their interactions that could case a trip. They predict if future activities at the plant would cause a reactor trip, provide a reactor trip warning system and aid in post-trip cause analysis. RAMSES is a general reliability engineering software package that uses concepts of artificial intelligence to provide unique capabilities on personal and mini-computers

  3. Residual heat estimation by using Cherenkov radiation in Tehran Research Reactor

    International Nuclear Information System (INIS)

    Arkani, M.; Gharib, M.

    2008-01-01

    An experiment is set up in Tehran 5 MW research reactor to observe Cherenkov radiation response during post-shutdown periods. An ordinary PC camera is used for this purpose. Theoretical estimation of the total power including decay heat and neutronic power is checked against detector response. A general agreement suggests that the same setup could equally serve as an independent channel for similar purposes in other reactors. This suggested that a similar setup based on present experience could be utilized in other reactors especially with the aim of fuel surveillance and monitoring.

  4. Residual heat estimation by using Cherenkov radiation in Tehran Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Arkani, M. [Department of Nuclear Engineering, Azad University, Tehran (Iran, Islamic Republic of); Gharib, M. [Tehran Research Reactor, Nuclear Science and Technology Research Institute (NSTRI), Tehran 14395-836 (Iran, Islamic Republic of)], E-mail: mgharib@aeoi.org.ir

    2008-11-11

    An experiment is set up in Tehran 5 MW research reactor to observe Cherenkov radiation response during post-shutdown periods. An ordinary PC camera is used for this purpose. Theoretical estimation of the total power including decay heat and neutronic power is checked against detector response. A general agreement suggests that the same setup could equally serve as an independent channel for similar purposes in other reactors. This suggested that a similar setup based on present experience could be utilized in other reactors especially with the aim of fuel surveillance and monitoring.

  5. Reactivity requirements and safety systems for heavy water reactors

    International Nuclear Information System (INIS)

    Kati, S.L.; Rustagi, R.S.

    1977-01-01

    The natural uranium fuelled pressurised heavy water reactors are currently being installed in India. In the design of nuclear reactors, adequate attention has to be given to the safety systems. In recent years, several design modifications having bearing on safety, in the reactor processes, protective and containment systems have been made. These have resulted either from new trends in safety and reliability standards or as a result of feed-back from operating reactors of this type. The significant areas of modifications that have been introduced in the design of Indian PHWR's are: sophisticated theoretical modelling of reactor accidents, reactivity control, two independent fast acting systems, full double containment and improved post-accident depressurisation and building clean-up. This paper brings out the evolution of design of safety systems for heavy water reactors. A short review of safety systems which have been used in different heavy water reactors, of varying sizes, has been made. In particular, the safety systems selected for the latest 235 MWe twin reactor unit station in Narora, in Northern India, have been discussed in detail. Research and Development efforts made in this connection are discussed. The experience of design and operation of the systems in Rajasthan and Kalpakkam reactors has also been outlined

  6. Effects of Physico-Chemical Post-Treatments on the Semi-Continuous Anaerobic Digestion of Sewage Sludge

    Directory of Open Access Journals (Sweden)

    Xinbo Tian

    2017-07-01

    Full Text Available Sludge production in wastewater treatment plants is increasing worldwide due to the increasing population. This work investigated the effects of ultrasonic (ULS, ultrasonic-ozone (ULS-Ozone and ultrasonic + alkaline (ULS+ALK post-treatments on the anaerobic digestion of sewage sludge in semi-continuous anaerobic reactors. Three conditions were tested with different hydraulic retention times (HRT, 10 or 20 days and sludge recycle ratios (R = QR/Qin (%: 50 or 100%. Biogas yield increased by 17.8% when ULS+ALK post-treatment was applied to the effluent of a reactor operating at 20 days HRT and at a 100% recycle ratio. Operation at 10 days HRT also improved the biogas yield (277 mL CH4/g VSadded (VS: volatile solids versus 249 mL CH4/g VSadded in the control. The tested post-treatment methods showed 4–7% decrease in effluent VS. The post-treatment resulted in a decrease in the cellular ATP (Adenosine tri-phosphate concentration indicating stress imposed on microorganisms in the reactor. Nevertheless, this did not prevent higher biogas production. Based on the results, the post-treatment of digested sludge or treating the sludge between two digesters is an interesting alternative to pre-treatments.

  7. A PWR reactor downcomer modification for reduction of ECC bypass flow during LOCA

    International Nuclear Information System (INIS)

    Popov, N.; Bosevski, T.

    1986-01-01

    The ECC bypass phenomenon in the PWR reactor down-comer, which delays the reactor vessel refilling, after cold leg large break LOCA accident, has been subject of analysis in this paper. In the paper, a particular construction modification of the reactor down-comer has been suggested by inserting vertical ribs, aimed to intensify the reactor ECC refilling following the LOCA accident, and to advance the thermal-hydraulics safety of post-accidental cooling of the PWR reactors. To verify the effectiveness of the suggested down-comer construction modification, some properly selected results, obtained by corresponding verified mathematical model, have been presented in this paper. (author)

  8. TREAT experiment M2 post-test examination

    International Nuclear Information System (INIS)

    Holland, J.W.; Teske, G.M.; Florek, J.C.

    1986-01-01

    Transient Reactor Test (TREAT) Facility experiment M2 was performed to evaluate the transient behavior of metal-alloy fuel under accident conditions to investigate the inherent safety features of the fuel in integral fast reactor (IFR) system designs. Objectives were to obtain early information on the key fuel behavior characteristics at transient overpower (TOP) conditions in metal-fueled fast reactors; namely, margin to cladding breach and extent of axial self-extrusion of fuel within intact cladding. The onset of cladding breaching depends on fuel/cladding eutectic formation, as well as cladding pressurization and melting. Driving forces for fuel extrusion are fission gas, liquid sodium, and volatile fission products trapped within the fuel matrix. The post-test examination provided data essential for correctly modeling fuel behavior in accident codes

  9. Socioeconomic consequences of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Tawil, J.J.; Callaway, J.W.; Coles, B.L.; Cronin, F.J.; Currie, J.W.; Imhoff, K.L.; Lewis, P.M.; Nesse, R.J.; Strenge, D.L.

    1984-06-01

    This report identifies and characterizes the off-site socioeconomic consequences that would likely result from a severe radiological accident at a nuclear power plant. The types of impacts that are addressed include economic impacts, health impacts, social/psychological impacts and institutional impacts. These impacts are identified for each of several phases of a reactor accident - from the warning phase through the post-resettlement phase. The relative importance of the impact during each accident phase and the degree to which the impact can be predicted are indicated. The report also examines the methods that are currently used for assessing nuclear reactor accidents, including development of accident scenarios and the estimating of socioeconomic accident consequences with various models. Finally, a critical evaluation is made regarding the use of impact analyses in estimating the contribution of socioeconomic consequences to nuclear accident reactor accident risk. 116 references, 7 figures, 15 tables

  10. Tritium release from lithium silicate and lithium aluminate, in-reactor and out-of-reactor

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.

    1976-09-01

    Studies were conducted to determine the generation and evolution of tritium and helium in lithium aluminate (LiAlO/sub 2/) and lithium silicate (Li/sub 2/SiO/sub 3/) by the reaction: Li/sup 6/ + n ..-->.. /sup 4/He + T. Targets were irradiated 4.4 days in the K-West Reactor snout facility. (Silicate GVR* approximately 2.0 cc/cc; aluminate GVR approximately 1.4 cc/cc.) Gas release in-reactor was determined by post-irradiation drilling experiments on aluminum ampoules containing silicate and aluminate targets. In-reactor tritium release (at approximately 100/sup 0/C) was found to decrease linearly with increasing target density. Tritium released in-reactor was primarily in the noncondensible form (HT and T/sub 2/), while in laboratory extractions (300-1300/sup 0/C), the tritium appeared primarily in the condensible form (HTO and T/sub 2/O). Concentrations of HT (and presumably HTO) were relatively high, indicating moisture pickup in canning operations or by inleakage of moisture after the capsule was welded. Impurities in extracted gases included H/sub 2/O, CO/sub 2/, CO, O/sub 2/, H/sub 2/, NO, SO/sub 2/, SiF/sub 4/ and traces of hydrocarbons.

  11. Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for Kijang research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Man; Tahk, Young Wook; Jeong, Yong Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); and others

    2017-08-15

    The construction project of the Kijang research reactor (KJRR), which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm{sup 3}, was selected to achieve higher fuel efficiency and performance than are possible when using U{sub 3}Si{sub 2}/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1)], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm{sup 3}), were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination) results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

  12. Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for kijang research reactor

    Directory of Open Access Journals (Sweden)

    Jong Man Park

    2017-08-01

    Full Text Available The construction project of the Kijang research reactor (KJRR, which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm3, was selected to achieve higher fuel efficiency and performance than are possible when using U3Si2/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm3, were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

  13. Estudio preliminar para el tratamiento de lixiviados en un reactor de biodiscos

    OpenAIRE

    Ordóñez Losada, Paola Jimena; Betancur Pérez, Alonso

    2003-01-01

    El presente trabajo hace parte de un proyecto de investigación de la Universidad Nacional de Colombia Sede Manizales y EMAS (Empresa Metropolitana de Aseo S.A. E.S.P) para encontrar la mejor alternativa para el tratamiento de los lixiviados del relleno sanitario “La Esmeralda” de la ciudad de Manizales, con el fin de cumplir la legislación ambiental vigente sobre vertimientos líquidos industriales a las aguas superficiales. Se analizó en forma preliminar la aplicación de la tecnología biodisc...

  14. De redes sociales recíprocas a grupos de acción para el intercambio de mercado : la “privatización espontánea” en la Hungría post-comunista

    OpenAIRE

    Lomnitz, Larissa

    2011-01-01

    Siguiendo el trabajo previo sobre la importancia que han tenido las redes sociales para la supervivencia económica y social del funcionariado de clase media latinoamericano y soviético, este artículo explora el papel de las redes sociales (las conexiones) en el proceso de privatización y liberalización del mercado en la Hungría post-comunista. Nos basamos en estudios académicos precedentes y en trabajo de campo desarrollado durante varios meses en Budapest para mostrar que las redes sociales ...

  15. Reference ZrH reactor power system for NASA space station post-operational reentry analysis

    International Nuclear Information System (INIS)

    Elliott, R.D.

    1970-01-01

    The flight dynamic and heating of a spent ZrH reactor power system returning from orbit at the end of its useful life are analyzed. The results of this analysis indicate that the reactor with a large portion of the lithium shield still surrounding it will impact the earth at a velocity of from 660 to 820 ft/sec, depending upon whether it tumbles or becomes stabilized during the latter part of its trajectory. (U.S.)

  16. Modelo de dimensionamiento del servicio web hosting dirigido a proveedores de la pequeña y mediana empresa colombiana

    Directory of Open Access Journals (Sweden)

    Olga Lucía Ramírez Calero

    2010-09-01

    Full Text Available La administración y dimensionamiento de recursos de infraestructura de un servicio web hosting, que consiste en proveer alojamiento para que una página web funcione correctamente [21], se realiza con un alto costo en servidores de elevado tamaño y disponibilidad, mediante el monitoreo de indicadores del nivel de ocupación de la capacidad que establecen reactivamente necesidades de modificación de la misma, o de otra forma, mediante una infraestructura redundante que significa incrementar el número de componentes para garantizar la disponibilidad del servicio. En un servicio dirigido a pequeñas y medianas empresas, se requiere cumplir ciertos niveles de disponibilidad a bajo costo. El presente artículo contiene los resultados de una investigación acerca del diseño, desarrollo y validación de un modelo de optimización de capacidades para el dimensionamiento y selección de la mejor combinación de recursos de infraestructura  (servidor, disco duro, memoria RAM y procesador en el servicio web hosting dirigido a PYMES colombianas.

  17. AGR-1 Post Irradiation Examination Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    The post-irradiation examination (PIE) of the Advanced Gas Reactor (AGR)-1 experiment was a multi-year, collaborative effort between Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) to study the performance of UCO (uranium carbide, uranium oxide) tristructural isotropic (TRISO) coated particle fuel fabricated in the U.S. and irradiated at the Advanced Test Reactor at INL to a peak burnup of 19.6% fissions per initial metal atom. This work involved a broad array of experiments and analyses to evaluate the level of fission product retention by the fuel particles and compacts (both during irradiation and during post-irradiation heating tests to simulate reactor accident conditions), investigate the kernel and coating layer morphology evolution and the causes of coating failure, and explore the migration of fission products through the coating layers. The results have generally confirmed the excellent performance of the AGR-1 fuel, first indicated during the irradiation by the observation of zero TRISO coated particle failures out of 298,000 particles in the experiment. Overall release of fission products was determined by PIE to have been relatively low during the irradiation. A significant finding was the extremely low levels of cesium released through intact coatings. This was true both during the irradiation and during post-irradiation heating tests to temperatures as high as 1800°C. Post-irradiation safety test fuel performance was generally excellent. Silver release from the particles and compacts during irradiation was often very high. Extensive microanalysis of fuel particles was performed after irradiation and after high-temperature safety testing. The results of particle microanalysis indicate that the UCO fuel is effective at controlling the oxygen partial pressure within the particle and limiting kernel migration. Post-irradiation examination has provided the final body of data that speaks to the quality of the AGR-1 fuel, building

  18. Completely automated nuclear reactors for long-term operation

    International Nuclear Information System (INIS)

    Teller, E.; Ishikawa, M.; Wood, L.

    1996-01-01

    The authors discuss new types of nuclear fission reactors optimized for the generation of high-temperature heat for exceedingly safe, economic, and long-duration electricity production in large, long-lived central power stations. These reactors are quite different in design, implementation and operation from conventional light-water-cooled and -moderated reactors (LWRs) currently in widespread use, which were scaled-up from submarine nuclear propulsion reactors. They feature an inexpensive initial fuel loading which lasts the entire 30-year design life of the power-plant. The reactor contains a core comprised of a nuclear ignitor and a nuclear burn-wave propagating region comprised of natural thorium or uranium, a pressure shell for coolant transport purposes, and automatic emergency heat-dumping means to obviate concerns regarding loss-of-coolant accidents during the plant's operational and post-operational life. These reactors are proposed to be situated in suitable environments at ∼100 meter depths underground, and their operation is completely automatic, with no moving parts and no human access during or after its operational lifetime, in order to avoid both error and misuse. The power plant's heat engine and electrical generator subsystems are located above-ground

  19. Study of indicators aggregation techniques for the selection of a new nuclear reactor for Mexico; Estudio de tecnicas de agregacion de indicadores para la seleccion de un nuevo reactor nuclear para Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Barragan M, A.M.; Martin del Campo M, C. [UNAM, Facultad de Ingenieria, 04510 Mexico D.F. (Mexico)]. e-mail: ale_bar_m@yahoo.com.mx

    2007-07-01

    A study on several aggregation techniques that can be used as multi criteria analysis methods, like important part of the methodology developed for the selection of a nuclear reactor for Mexico is described. In an arbitrary way three reactors were selected to be compared, these they are the AP1000 (Advance Passive from 1000 MWe), the PBMR (Pebble Bed Modular Reactor) and the GT-MHR (Gas Turbine Modular Helium). The evaluation approaches were classified in three categories: Economic, Socio-political and of safety and environment. In each category they were defined the more important evaluation indicators and then it was built a matrix with those values of each reactor. The four studied aggregation methods are described: Normalization, Linear deliberation, Fuzzy Logic and AHP (Analytic Hierarchy Process). The well-known aggregation mechanisms are those that are obtained of the lineal deliberation and of the normalization, which have demonstrated to give good results before the simplicity of their use. The fuzzy logic has the advantage that it allows to manage qualitative and quantitative information simultaneously without the aggregation problems that are presented since in a conventional system the semantic pattern on that is based, it is provided by the theory of the diffuse groups that has demonstrated in other areas of the knowledge a better approach to the reality, when admitting that the nature has shades and that the decisions take in function of a wide range of possibilities and of approaches in contradictory occasions or in conflict, all equally worth. The Analytic Hierarchical Process (AHP) that consists in formalizing the intuitive understanding of a multi criteria complex problem, by means of the construction of a hierarchical model that allows the decision agent to structure the problem in visual form, giving him the form of a hierarchy of attributes (global objective of the problem, approaches and alternative). Finally, using the matrix of initiators

  20. Engineering feasibility of tight aspect ratio Tokamak (spherical torus) reactors

    International Nuclear Information System (INIS)

    Peng, Y-K.M.; Hicks, J.B.

    1990-01-01

    Engineering solutions are identified and analyzed for key high-power-density components of tight aspect ratio tokamak reactors (spherical torus reactors). The potentially extreme divertor heat loads can be reduced to about 3 MW/m 2 in expanded divertors using coils inside the demountable toroidal field coils. Given the long and narrow divertor channels, gaseous divertor targets become possible, which eliminate sputtering and increase the divertor life. The unshielded centre conductor post (CCP) of the toroidal field coil can be made of a single dispersion strengthened copper conductor cooled by high-velocity pressurized water to maintain acceptable copper temperature and strength. Damage and activation of the CCP at a neutron fluence of 10 MW-a/m 2 are also tolerable. Annual replacement of the centre post, the divertor assemblies and the blanket can be accomplished with vertical access for all torus components, which are modularized to reduce size and weight. The technical requirements of these solutions are shown to be comparable with, if not less demanding than, those estimated for conventional tokamak reactors. (author)

  1. After-operating properties of nuclear reactor vessel materials of Lenin atomic ice breaker and prospective of reactor vessels radiation life prolongation

    International Nuclear Information System (INIS)

    Platonov, P.A.; Shtrombakh, Ya.I.; Amaev, A.D.; Krasikov, E.A.; Korolev, Yu.N.; Zabusov, O.O.; Glushakov, G.M.

    2001-01-01

    A post-operational state of the icebreaker Lenin reactor vessel metal is investigated. It is shown that a base metal of the icebreaker Lenin reactor vessel is of high quality as by an initial value of critical temperature of embrittlement, so by its radiation resistance. The weld metal possesses a sufficient radiation resistance but has an insufficient initial ductile-brittle transition temperature (approximately 63 Deg C). It is necessary to note that the final stage of operation for nuclear steam-generating plant should be carried out at the coolant temperature as high as possible [ru

  2. Administración del recurso hídrico para consumo humano en Costa Rica y su consecuencia en la mortalidad infantil post-neonatal

    Directory of Open Access Journals (Sweden)

    Edwin Vega Araya

    2008-06-01

    Full Text Available Si se está pensando en reducir la mortalidad infantil, es útil dividir la mortalidad según el tiempo en que ocurre. La mortalidad perinatal y neonatal (en los primeros días de vida hasta los 27 días de vida y la mortalidad post-neonatal (entre 28 y 365 días de vida. El presente estudio se refiere a este último periodo de deceso, la tasa de mortalidad infantil post-neonatal (TMI post-neonatal, que si bien tiene un peso relativo menor, no deja de ser importante. Primeramente se analiza la influencia de otros factores, como los factores sociodemográficos a través del estudio de la relación entre el Índice de Desarrollo Social (IDS y la TMI post-neonatal. Luego se establece cuáles son los entes que proveen el servicio de agua potable en los diferentes distritos del país, y, dadas sus diferentes características, se analiza y verifica la hipótesis de que el ente proveedor (o administrador es un factor que explica la variación de la TMI post-neonatal entre distritos. Para reducir la TMI post-neonatal se debe obrar en la calidad del agua potable. La primera sugerencia es que en cada distrito debe haber un claro responsable y proveedor del agua potable. A juzgar por el éxito de ICAA y ESPH, parecen más exitosos los entes administradores basados en el control (en términos de definición de tarifas y control de calidad por parte de ARESEP, y no por Concejos Municipales que son más susceptibles a aplazar medidas como la colocación de medidores, la ejecución de ciertas inversiones, etc.

  3. Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, October--December 1975

    International Nuclear Information System (INIS)

    1976-05-01

    Light water reactor safety activities performed during October--December 1975 are reported. The blowdown heat transfer tests series of the Semiscale Mod-1 test program was completed. In the LOFT Program, preparations were made for nonnuclear testing. The Thermal Fuels Behavior Program completed a power-cooling-mismatch test and an irradiation effects test on PWR-type fuel rods. Model development and verification efforts of the Reactor Behavior Program included developing new analysis models for the RELAP4 computer code, subroutines for the FRAP-S and FRAP-T codes, and new models for predicting reactor fuel restructuring and zircaloy cladding behavior; an analysis of post-CHF fuel behavior was made using FRAP-T

  4. Study of thermophysical and thermohydraulic properties of sodium for fast sodium cooled reactors; Estudio de las propiedades termofisicas y termohidraulicas del sodio para reactores rapidos enfriados por sodio

    Energy Technology Data Exchange (ETDEWEB)

    Vega R, A. K.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: a.karen.vr@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The importance of liquid sodium lies in its use as a coolant for fast reactors, but why should liquid metal be used as a coolant instead of water? Water is difficult to use as a coolant for a fast nuclear reactor because its acts as a neutron moderator, that is, stop the fast neutrons and converts them to thermal neutrons. Nuclear reactors such as the Pressurized Water Reactor or the Boiling Water Reactor are thermal reactors, which mean they need thermal neutrons for their operation. However, is necessary for fast reactors to conserve as much fast neutrons, so that the liquid metal coolants that do have this capability are implemented. Sodium does not need to be pressurized, its low melting point and its high boiling point, higher than the operating temperature of the reactor, make it an adequate coolant, also has a high thermal conductivity, which is necessary to transfer thermal energy and its viscosity is close to that of the water, which indicates that is an easily transportable liquid and does not corrode the steel parts of the reactor. This paper presents a brief state of the art of the rapid nuclear reactors that operated and currently operate, as well as projects in the door in some countries; types of nuclear reactors which are cooled by liquid sodium and their operation; the mathematical models for obtaining the properties of liquid sodium in a range of 393 to 1673 Kelvin degrees and a pressure atmosphere. Finally a program is presented in FORTRAN named Thermo-Sodium for the calculation of the properties, which requires as input data the Kelvin temperature in which the liquid sodium is found and provides at the user the thermo-physical and thermo-hydraulic properties for that data temperature. Additional to this the user is asked the Reynolds number and the hydraulic diameter in case of knowing them, and in this way the program will provide the value of the convective coefficient and that of the dimensionless numbers: Nusselt, Prandtl and Peclet. (Author)

  5. A personal computer based console monitor for a TRIGA reactor

    International Nuclear Information System (INIS)

    Rieke, Phillip E.; Hood, William E.; Razvi, Junaid

    1990-01-01

    Numerous improvements have been made to the Mark F facility to provide a minimum reactor down time, giving a high reactor availability. A program was undertaken to enhance the monitoring capabilities of the instrumentation and control system on this reactor. To that end, a personal computer based console monitoring system has been developed, installed in the control room and is operational to provide real-time monitoring and display of a variety of reactor operating parameters. This system is based on commercially available hardware and an applications software package developed internally at the GA facility. It has (a) assisted the operator in controlling reactor parameters to maintain the high degree of power stability required during extended runs with thermionic devices in-core, and (b) provided data trending and archiving capabilities on all monitored channels to allow a post-mortem analysis to be performed on any of the monitored parameters

  6. Handbook of materials testing reactors and ancillary hot laboratories in the European Community

    International Nuclear Information System (INIS)

    1977-01-01

    The purpose of this Handbook is to make available to those interested in 'in-pile' irradiation experiments important data on Materials Testing Reactors in operation in the European Community. Only thermal reactors having a power output of more than 5 MW(th) are taken into consideration. In particular, detailed technical information is given on the experimental irradiation facilities of the reactors, their specialized irradiation devices (loops and instrumented capsules), and the associated hot cell facilities for post-irradiation examination of samples

  7. Análisis de señales de medidas mecánicas para el mantenimiento predictivo avanzado

    OpenAIRE

    Montalvo Martín, Cristina

    2012-01-01

    Para contribuir al diseño de un mantenimiento mecánico avanzado, se utilizan en esta tesis las técnicas de análisis de ruido para monitorizar las vibraciones de los internos de un reactor PWR y para vigilar la respuesta dinámica de los sensores de presión capacitivos tipo Rosemount ampliamente utilizados en la industria, sobre todo en las plantas nucleares. Para el primer caso, se han ajustado mediante un método no lineal de Breit- Wigner los espectros de resonancias obtenidos por medio de la...

  8. Análisis de señales de medidas mecánicas para el mantenimiento predictivo avanzado

    OpenAIRE

    Montalvo Martín, Cristina

    2010-01-01

    Para contribuir al diseño de un mantenimiento mecánico avanzado, se utilizan en esta tesis las técnicas de análisis de ruido para monitorizar las vibraciones de los internos de un reactor PWR y para vigilar la respuesta dinámica de los sensores de presión capacitivos tipo Rosemount ampliamente utilizados en la industria, sobre todo en las plantas nucleares. Para el primer caso, se han ajustado mediante un método no lineal de Breit- Wigner los espectros de resonancias obtenidos por medio de la...

  9. Post-irradiation data analysis for NRC/PNL Halden assembly IFA-431

    International Nuclear Information System (INIS)

    Nealley, C.; Lanning, D.D.; Cunningham, M.E.; Hann, C.R.

    1979-10-01

    Results are presented for the post irradiation examination performed on IFA-431, which was a 6-rod test fuel assembly irradiated in Halden Reactor, Norway, under sponsorship of the Nuclear Regulatory Commission. The irradiation conditions included: peak powers of 33 kW/m; coolant pressure and temperature of 3.3 MPa and 240 0 C, respectively; and peak burnup of 4300 MWd/MTM. IFA-431 included instrumented rods of basic boiling water reactor design, with variations in fill gas composition, gap size, and UO 2 fuel type. The irradiation was designed to measure the effect of these variations upon fuel rod thermal and mechanical performance. The post irradiation examination assessed the permanent changes to the rods, including induced radioactivity, cladding deformation, fission gas release, and fuel densification

  10. Reviviendo la consulta post-mortem.

    OpenAIRE

    Armando Cortés

    2009-01-01

    Por estos días se inaugura el “Centro de consulta post-mortem del Hospital Universitario del Valle”, una denominación más apropiada para la autopsia «ver por sí mismo» o cualquiera de sus sinónimos necropsia, examen post-mortem, necroscopia, o tanatopsia; todos ellos no aceptados y condicionados por factores culturales, sociales o religiosos. Estos términos han alcanzado una connotación claramente negativa en el ambiente médico y en el público general. Quizás, el mejor término sea «consulta p...

  11. Aspects of nuclear reactor safety

    International Nuclear Information System (INIS)

    Hardt, P. von der; Rottger, H.

    1980-01-01

    The Colloquium on 'Irradiation Tests for Reactor Safety Programmes' has been organised by JRC Petten in order to determine the present state of technology in the field. The role of research and test reactors for studies of structural material and fuel elements under transient and off-normal conditions was to be explained. The Colloquium has been attended by 110 participants from outside and inside Europe. 27 papers were presented covering the major ongoing projects in Japan, the United States, and in Europe, and elaborating in particular: - design rationale and layout of safety irradiation experiments; - design, manufacture, and performance of irradiation equipment with particular attention to generation and control of transient conditions, fast response in-pile instrumentation and its out-of-pile data retrieval; - post-irradiation evaluation; - results and analytical support

  12. 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.

  13. Compact light-emitting diode optical fiber immobilized TiO2 reactor for photocatalytic water treatment.

    Science.gov (United States)

    O'Neal Tugaoen, Heather; Garcia-Segura, Sergi; Hristovski, Kiril; Westerhoff, Paul

    2018-02-01

    A key barrier to implementing photocatalysis is delivering light to photocatalysts that are in contact with aqueous pollutants. Slurry photocatalyst systems suffer from poor light penetration and require post-treatment to separate the catalyst. The alternative is to deposit photocatalysts on fixed films and deliver light onto the surface or the backside of the attached catalysts. In this study, TiO 2 -coated quartz optical fibers were coupled to light emitting diodes (OF/LED) to improve in situ light delivery. Design factors and mechanisms studied for OF/LEDs in a flow-through reactor included: (i) the influence of number of LED sources coupled to fibers and (ii) the use of multiple optical fibers bundled to a single LED. The light delivery mechanism from the optical fibers into the TiO 2 coatings is thoroughly discussed. To demonstrate influence of design variables, experiments were conducted in the reactor using the chlorinated pollutant para-chlorobenzoic acid (pCBA). From the degradation kinetics of pCBA, the quantum efficiencies (Φ) of oxidation and electrical energies per order (E EO ) were determined. The use of TiO 2 coated optical fiber bundles reduced the energy requirements to deliver photons and increased available surface area, which improved Φ and enhanced oxidative pollutant removal performance (E EO ). Copyright © 2017 Elsevier B.V. All rights reserved.

  14. Termoterapia para o controle de patógenos em pós-colheita em frutos da cajazeira = Thermotherapy for post harvest pathogens on Spondias fruits

    Directory of Open Access Journals (Sweden)

    Carlos Henrique de Brito

    2008-01-01

    Full Text Available O tratamento térmico, principalmente água quente, é método alternativo que tem sido utilizado para o controle de doenças e infestações de insetos em frutos póscolheita. O presente trabalho teve como objetivo determinar a combinação de tempo e temperatura adequada para o controle de fungos de pós-colheita em frutos de cajazeira em atmosfera ambiente. No primeiro tratamento, os frutos foram imersos em água quente e no segundo foram expostos ao vapor a 50°C por 0, 10, 20, 30 e 40 minutos para diferentes lotes de frutos. Foram retiradas de cada fruto/tratamento quatro secções, as quais foram incubadas em placas de Petri com BDA, sendo realizadas as avaliações da incidência de fitopatógenos após 7 dias de incubação. Os resultados obtidos demonstraram uma maior incidência de Rhizophus sp. nos tratamentos avaliados e redução de Aspergillus sp. e Fusariumsp., cujo comportamento foi influenciado pelo tratamento termoterápico, podendo ser indicado os tratamentos vapor e banho-maria a 50ºC a partir de 20 minutos como método alternativo no controle pós-colheita de Aspergillus sp. e Fusarium sp. em frutos da cajazeira.Thermal treatment, mainly hot water, is an alternative method that has been used for diseases and pests infestation in post harvest fruits. The present work aimed to determine a combination of correct time x temperature for post harvest fungus control on Spondias fruits. For the first treatment, fruits were dipped on hot water and, for the second, on hot air, both with 50°C for 0, 10, 20, 30 e 40 minutes for different fruit groups. Four pieces were sectioned from each fruit, per treatment, and incubated in Petri dishes with BDA,being evaluated for fungus incidence after seven days incubation. Obtained results showed higher incidence of Rhizopus sp. on the evaluated treatments, and a reduction of Aspergillus sp. and Fusarium sp., while behaviour was influenced by thermotherapy, indicating air and hot water at 50º

  15. La digestión anaerobia y los reactores UASB. Generalidades

    OpenAIRE

    Yaniris Lorenzo; Ma. Cristina Obaya

    2006-01-01

    Se muestran las generalidades de los reactores, se da a conocer su concepto, se enumeran las aguas residuales que pueden ser tratadas en los mismos, se comentan los parámetros a tener en cuenta para que funcione adecuadamente y se enumeran las ventajas y desventajas de este proceso, así como su aplicabilidad.

  16. The economics of Magnox reactors

    International Nuclear Information System (INIS)

    Watts, P.E.

    1982-01-01

    The CEGB regularly publishes figures for generation costs per kilowatt hour (kWh) at Magnox and several coal and oil stations. These have been criticized by Select Committees and most recently by Professor Jeffery for being expressed in historic cost accounting terms rather than in constant price terms. How the economic performance of Magnox reactors can best be judged is discussed. Headings are: post audit of past investment; future investment choice; inflation corrected assets. (author)

  17. Meso-scale modeling of irradiated concrete in test reactor

    International Nuclear Information System (INIS)

    Giorla, A.; Vaitová, M.; Le Pape, Y.; Štemberk, P.

    2015-01-01

    Highlights: • A meso-scale finite element model for irradiated concrete is developed. • Neutron radiation-induced volumetric expansion is a predominant degradation mode. • Confrontation with expansion and damage obtained from experiments is successful. • Effects of paste shrinkage, creep and ductility are discussed. - Abstract: A numerical model accounting for the effects of neutron irradiation on concrete at the mesoscale is detailed in this paper. Irradiation experiments in test reactor (Elleuch et al., 1972), i.e., in accelerated conditions, are simulated. Concrete is considered as a two-phase material made of elastic inclusions (aggregate) subjected to thermal and irradiation-induced swelling and embedded in a cementitious matrix subjected to shrinkage and thermal expansion. The role of the hardened cement paste in the post-peak regime (brittle-ductile transition with decreasing loading rate), and creep effects are investigated. Radiation-induced volumetric expansion (RIVE) of the aggregate cause the development and propagation of damage around the aggregate which further develops in bridging cracks across the hardened cement paste between the individual aggregate particles. The development of damage is aggravated when shrinkage occurs simultaneously with RIVE during the irradiation experiment. The post-irradiation expansion derived from the simulation is well correlated with the experimental data and, the obtained damage levels are fully consistent with previous estimations based on a micromechanical interpretation of the experimental post-irradiation elastic properties (Le Pape et al., 2015). The proposed modeling opens new perspectives for the interpretation of test reactor experiments in regards to the actual operation of light water reactors.

  18. Meso-scale modeling of irradiated concrete in test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giorla, A. [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Vaitová, M. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic); Le Pape, Y., E-mail: lepapeym@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Štemberk, P. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic)

    2015-12-15

    Highlights: • A meso-scale finite element model for irradiated concrete is developed. • Neutron radiation-induced volumetric expansion is a predominant degradation mode. • Confrontation with expansion and damage obtained from experiments is successful. • Effects of paste shrinkage, creep and ductility are discussed. - Abstract: A numerical model accounting for the effects of neutron irradiation on concrete at the mesoscale is detailed in this paper. Irradiation experiments in test reactor (Elleuch et al., 1972), i.e., in accelerated conditions, are simulated. Concrete is considered as a two-phase material made of elastic inclusions (aggregate) subjected to thermal and irradiation-induced swelling and embedded in a cementitious matrix subjected to shrinkage and thermal expansion. The role of the hardened cement paste in the post-peak regime (brittle-ductile transition with decreasing loading rate), and creep effects are investigated. Radiation-induced volumetric expansion (RIVE) of the aggregate cause the development and propagation of damage around the aggregate which further develops in bridging cracks across the hardened cement paste between the individual aggregate particles. The development of damage is aggravated when shrinkage occurs simultaneously with RIVE during the irradiation experiment. The post-irradiation expansion derived from the simulation is well correlated with the experimental data and, the obtained damage levels are fully consistent with previous estimations based on a micromechanical interpretation of the experimental post-irradiation elastic properties (Le Pape et al., 2015). The proposed modeling opens new perspectives for the interpretation of test reactor experiments in regards to the actual operation of light water reactors.

  19. Recent developments in post-irradiation examination techniques for water reactor fuel. Proceedings of a technical committee meeting held in Cadarache, France, 17-21 October 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    At the invitation of the Government of France, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) the IAEA convened a Technical Committee meeting from 14 to 21 October 1994 in Cadarache to discuss recent technical advances and improvements in the field of post-irradiation examination (PIE) of fuel used in nuclear power plants. Fifty participants representing 14 countries attended the meeting and 30 papers were presented and discussed during five technical sessions. Working Groups composed of the session chairmen and authors of papers prepared summaries of each session including conclusions and recommendations for future work. Refs, figs and tabs.

  20. Recent developments in post-irradiation examination techniques for water reactor fuel. Proceedings of a technical committee meeting held in Cadarache, France, 17-21 October 1994

    International Nuclear Information System (INIS)

    1995-09-01

    At the invitation of the Government of France, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) the IAEA convened a Technical Committee meeting from 14 to 21 October 1994 in Cadarache to discuss recent technical advances and improvements in the field of post-irradiation examination (PIE) of fuel used in nuclear power plants. Fifty participants representing 14 countries attended the meeting and 30 papers were presented and discussed during five technical sessions. Working Groups composed of the session chairmen and authors of papers prepared summaries of each session including conclusions and recommendations for future work. Refs, figs and tabs

  1. AECL hot-cell facilities and post-irradiation examination services

    International Nuclear Information System (INIS)

    Schankula, M.H.; Plaice, E.L.; Woodworth, L.G.

    1998-04-01

    This paper presents an overview of the post-irradiation examination (PIE) services available at AECL's hot-cell facilities (HCF). The HCFs are used primarily to provide PIE support for operating CANDU power reactors in Canada and abroad, and for the examination of experimental fuel bundles and core components irradiated in research reactors at the Chalk River Laboratories (CRL) and off-shore. A variety of examinations and analyses are performed ranging from non-destructive visual and dimensional inspections to detailed optical and scanning electron microscopic examinations. Several hot cells are dedicated to mechanical property testing of structural materials and to determine the fitness-for-service of reactor core components. Facility upgrades and the development of innovative examination techniques continue to improve AECL's PIE capabilities. (author)

  2. AECL hot-cell facilities and post-irradiation examination services

    International Nuclear Information System (INIS)

    Schankula, M.H.; Plaice, E.L.; Woodworth, L.G.

    1995-01-01

    This paper presents an overview of the post-irradiation examination (PIE) services available at AECL's hot-cell facilities (HCF). The HCFs are used primarily to provide PIE support for operating CANDU power reactors in Canada and abroad, and for the examination of experimental fuel bundles and core components irradiated in research reactors at the Chalk River Laboratories (CRL) and off-shore. A variety of examinations and analysis are performed ranging from non-destructive visual and dimensional inspections to detailed optical and scanning electron microscopic examinations. Several hot cells are dedicated to mechanical property testing of structural materials and to determine the fitness-for-service of reactor core components. Facility upgrades and the development of innovative examination techniques continue to improve AECL's PIE capabilities. (author)

  3. Power Nuclear Reactors: technology and innovation for development in future; Centrales Nucleares de Potencia: tecnologias actuales e innovaciones para el futuro

    Energy Technology Data Exchange (ETDEWEB)

    Suarez Antola, R [Universidad Catolica del Uruguay, Montevideo(Uruguay); Ministerio de Industria Energia y Minerria, Montevideo(Uruguay)

    2009-07-01

    The conference is about some historicals task of the fission technology as well as many types of Nuclear Reactors. Enrichment of fuel, wastes, research reactors and power reactors, a brief advertisment about Uruguay electric siystem and power generation, energetic worldwide, proliferation, safety reactors, incidents, accidents, Three-Mile Island accident, Chernobil accident, damages, risks, classification and description of Power reactors steam generation, nuclear reactor cooling systems, future view.

  4. The Jules Horowitz Reactor project, a driver for revival of the research reactor community

    International Nuclear Information System (INIS)

    Pere, P.; Cavailler, C.; Pascal, C.

    2010-01-01

    The first concrete of the nuclear island for the Jules Horowitz Reactor (JHR) was poured at the end of July 2009 and construction is ongoing. The JHR is the largest new platform for irradiation experiments supporting Generation II and III reactors, Generation IV technologies, and radioisotope production. This facility, composed of a unique grouping of workshops, hot cells and hot laboratories together with a first-rate MTR research reactor, will ensure that the process, from preparations for irradiation experiments through post-irradiation non-destructive examination, is completed expediently, efficiently and, of course, safely. In addition to the performance requirements to be met in terms of neutron fluxes on the samples (5x10 14 n.cm -2 /sec -1 E>1 MeV in core and 3,6x10 14 n.cm -2 /sec -1 E<0.625 eV in the reflector) and the JHR's considerable irradiation capabilities (more than 20 experiments and one-tenth of irradiation area for simultaneous radioisotope production), the JHR is the first MTR to be built since the end of the 1960s, making this an especially challenging project. The presentation will provide an overview of the reactor, hot cells and laboratories and an outline of the key milestones in the project schedule, including initial criticality in early 2014 and radioisotope production in 2015. This will be followed by a description of the project organization set up by the CEA as owner and future operator and AREVA TA as prime contractor and supplier of critical systems, and a discussion of project challenges, especially those dealing with the following items:accommodation of a broad experimental domain; involvement by international partners making in-kind contributions to the project; ? development of components critical to safety and performance; the revival of engineering of research reactors and experimental devices involving France's historical players in the field of research reactors, and; tools to carry out the project, including computer codes

  5. Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, October--December 1975

    Energy Technology Data Exchange (ETDEWEB)

    1976-05-01

    Light water reactor safety activities performed during October--December 1975 are reported. The blowdown heat transfer tests series of the Semiscale Mod-1 test program was completed. In the LOFT Program, preparations were made for nonnuclear testing. The Thermal Fuels Behavior Program completed a power-cooling-mismatch test and an irradiation effects test on PWR-type fuel rods. Model development and verification efforts of the Reactor Behavior Program included developing new analysis models for the RELAP4 computer code, subroutines for the FRAP-S and FRAP-T codes, and new models for predicting reactor fuel restructuring and zircaloy cladding behavior; an analysis of post-CHF fuel behavior was made using FRAP-T.

  6. Hydraulic shock damper for fuel assemblies of nuclear reactors

    International Nuclear Information System (INIS)

    Jabson, F.S.

    1978-01-01

    A typical embodiment of this invention provides a hydraulic mechanism for alleviating the effect of seismic forces and other stresses that are applied to a fuel assembly in a nuclear reactor. Illustratively, hollow guide posts potrude into a fuel assembly end fitting grid from biased spring pads. Plungers that move with the spring pads plug one end of each of the respective guide posts. Plates on the end fitting grid that have individual holes for fluid discharge partially plug the other ends of the respective guide posts, thereby providing a hydraulic means for absorbing the longitudinal component of seismic shocks and other anticipated forces. (Auth.)

  7. An Analysis Of Conjunctions In The Jakarta Post Editorials

    OpenAIRE

    Dewi, Ika Sari

    2010-01-01

    Skripsi ini berjudul An Analysis of Conjunctions in the Jakarta Post Editorials, yang membahas tentang penggunaan conjunctions dalam kalimat yang terdapat pada harian Jakarta Post. Penelitian ini mengambil 6 editorial sebagai bahan penelitiannya. Skripsi ini ditulis berdasarkan keingintahuan penulis tentang sejauh mana para editor menggunakan kata sambung dalam tulisan-tulisan yang mereka hasilkan. Dalam mengerjakan skripsi ini, penulis menerapkan studi kepustakaan yakni dengan mengumpulka...

  8. System for cooling the containment vessel of a nuclear reactor

    International Nuclear Information System (INIS)

    Costes, Didier.

    1982-01-01

    The invention concerns a post-accidental cooling system for a nuclear reactor containment vessel. This system includes in series a turbine fed by the moist air contained in the vessel, a condenser in which the air is dried and cooled, a compressor actuated by the turbine and a cooling exchanger. The cold water flowing through the condenser and in the exchanger is taken from a tank outside the vessel and injected by a pump actuated by the turbine. The application is for nuclear reactors under pressure [fr

  9. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  10. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    International Nuclear Information System (INIS)

    Shropshire, D.E.

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program's understanding of the cost drivers that will determine nuclear power's cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-irradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  11. Técnica del Poste Anatómico (Grandini: Caso clínico

    Directory of Open Access Journals (Sweden)

    Sergio Pignata Volpe

    Full Text Available Resumen Uno de los grandes desafíos de los postes prefabricados de fibra ha implicado, desde sus orígenes, mejorar su diseño buscando una mayor adaptación al conducto radicular. Esto no sólo importa a los efectos de que la capa de cemento sea lo más delgada posible, sino también porque el íntimo contacto entre poste y conducto radicular genera un mecanismo de retención por fricción, que es muy favorable en el desempeño del poste para evitar su descementado. Una de las formas de mejorar dicha adaptación es a través de la Técnica del Poste Anatómico desarrollada por Grandini. En este artículo se presenta una revisión bibliográfica del tema, describiendo la técnica para un caso clínico con contraindicación de anclaje colado por el alto compromiso de resistencia corono-radicular. Palabras clave: poste anatómico, postes prefabricados de fibra, cementos adhesivos, contracción de polimerización, Técnica del Perno Muñón

  12. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

    2007-06-15

    The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

  13. Remoção de matéria orgânica e sólidos suspensos por nova configuração de biofiltro aeróbio submerso no pós-tratamento de efluente de reator UASB Removal of organic matter and suspended solids by a new configuration of biological aerated filter in the post-treatment of UASB reactor effluent

    Directory of Open Access Journals (Sweden)

    Saulo Varela Della Giustina

    2010-09-01

    Full Text Available O pós-tratamento de efluentes de reatores anaeróbios é um processo necessário para o atendimento dos padrões de emissão. Os resultados aqui apresentados mostram a viabilidade de uso de uma nova configuração de biofiltro aeróbio submerso (BAS no pós-tratamento desses efluentes. Os BAS multiestágio apresentam uma câmara anaeróbia (V=12,6L, seguido de uma câmara aeróbia (V=30L e uma câmara anóxica (V=26,4L, todas em série (V total=70L. Neste estudo, foi analisada a remoção de sólidos suspensos (SS, DQO e DBO5. Foram utilizados três BAS multi-estágio preenchidos com três diferentes materiais-suporte: tampas e gargalos PET (165m²/m³, pedra britada n. 4 (50m²/m³ e anéis Pall 1,5'' (135m²/m³. Os reatores foram operados com valores de tempos de detenção hidráulicas (TDH de 4,1, 8,2 e 12,3 horas, e três taxas de aplicação superficial (TAS (21, 12 e 8m³/m².d. A associação dos reatores UASB+BAS possibilitou remoções de DQO total superiores a 90% para os BAS 1 e 3, e 85% para o BAS 2, sendo independente do TDH aplicado. A remoção de SS foi maior no BAS contendo anéis Pall, provavelmente devido ao maior índice de vazios desse material.The post-treatment of effluents from anaerobic reactors is normally a mandatory step to meet the emission standards. The results presented here show the feasibility of using a new configuration of biological aerated filter (BAF in the post-treatment of UASB reactors. The multi-stage BAF presents an anaerobic chamber (V=12.6L, followed by an aerobic chamber (V=30L and an anoxic chamber (V=26.4L, all in series (total V=70L. This study examined the removal of suspended solids (SS, COD and BOD5. Three multi-stage BAF filled with three different packing materials were used: lids and bottlenecks of PET bottles (165m²/m³, gravel n. 4 (50m²/m³ and Pall rings 1.5'' (135m²/m³. The reactors were operated with the values of hydraulic detention time (HDT of 4.1, 8.2 and 12.3 hours, and

  14. FOTOBIORREACTOR: HERRAMIENTA PARA CULTIVO DE CIANOBACTERIAS

    Directory of Open Access Journals (Sweden)

    Luis Guillermo Ramírez Mérida

    2014-05-01

    Full Text Available Las cianobactérias son organismos eficientes en la conversión de energía solar y producen una gran variedad de metabolitos. En la actualidad son el centro de atención para la producción de biocombustible, son usadas como biofertilizantes, control de contaminación ambiental y como fuente de nutrientes en alimentación humana y animal. Con el fin de proporcionar crecimiento y aprovechar el potencial de las cianobacterias, se requieren fotobiorreactores eficientes. Aunque se han propuesto muchos tipos de fotobiorreactores, no existe un reactor ideal, solo unos pocos pueden utilizarse para la producción de biomasa de cianobacterias. De hecho, la elección del fotobiorreactor más adecuado depende de la situación, ya que tanto las especies de algas disponibles y el destino final jugarán un papel importante. Uno de los principales factores que limita su aplicación práctica en cultivos de biomasa es la transferencia de masa. Por esto, entender el coeficiente de transferencia de masa en los fotobiorreactores es necesario para una operación eficiente del cultivo de biomasa en cianobacterias. En esta revisión, se discuten varios tipos de fotobiorreactores muy promisorios para la producción de biomasa de cianobacterias.

  15. Reactor inventory monitoring system for Angra-1 reactor; Sistema de monitoracao de inventario do reator para usina nuclear Angra I

    Energy Technology Data Exchange (ETDEWEB)

    S Neto, Joaquim A.; Silva, Marcos C.; Pinheiro, Ronaldo F.M. [Furnas Centrais Eletricas S.A., Rio de Janeiro, RJ (Brazil); Soares, Milton [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil); Martinez, Aquilino; Comerlato, Cesar A.; Oliveira, Eugenio A. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Lab. de Monitoracao de Processos

    1996-07-01

    This work describes the project of Reactor Inventory Monitoring System, which will be installed in Angra I Nuclear Power Plant. The inventory information is important to the operators take corrective actions in case of an incident that may cause a failure in the core cooling. (author)

  16. The Role of Exponential and PCTR Experiments at Hanford in the Design of Large Power Reactors; Roles Respectifs des Experiences Exponentielles et du Reacteur d'Etude des Constantes Physiques de Hanford dans les Etudes de Grands Reacteurs de Puissance; Znachenie ehksponentsial'nykh opytov i opytov na reaktore PCTR pri proektirovanii bol'shikh ehnergeticheskikh reaktorov v khehnforde; Papel de los Experimentos Exponenciales y del Reactor PCTR de Hanford en el Proyecto de Grandes Reactores de Potencia

    Energy Technology Data Exchange (ETDEWEB)

    Heineman, R. E. [General Electric Company, Richland, WA (United States)

    1964-02-15

    que ces installations permettent d'utiliser, en vue de faire face aux besoins de donnees experimentales de plus en plus diverses. Il faut avoir tous ces renseignements presents a l'esprit si l'on veut prevoir comment evolueront les besoins et les tendances dans l'emploi de ces installations pour les etudes de reacteurs de puissance. Le memoire decrit brievement le Reacteur d'etude des reseaux a haute temperature et indique comment on se propose de l'utiliser dans le cadre de cette evolution. (author) [Spanish] Desde hace casi 15 anos se vienen realizando en los laboratorios de Hanford mediciones exponenciales en reticulados de grafito* uranio. Aunque los resultados de dichos experimentos se emplearon para establecer los laplacianos de reactores de produccion, contribuyeron tambien a ampliar los conocimientos sobre la fisica de estos sistemas. Muy pronto se reconocio que la utilidad del experimento exponencial quedaba limitada por sus grandes dimensiones y por su escasa sensibilidad a pequenas perturbaciones localizadas del sistema. Por ello se comenzo a idear un experimento integral en un reactor que reduciria al minimo la cantidad de materiales necesarios para obtener datos significativos. A tal efecto, se construyo una instalacion critica perfeccionada de varias regiones, que se denomino PCTR (reactor para estudio de constantes fisicas). Este reactor se ha empleado para determinar las constantes fisicas de varios reactores de potencia. Ademas, ha servido como instalacion de uso general para medir secciones eficaces y para determinar los parametros diferenciales e integrales de fisica de los reactores correspondientes a diversos tipos de medios multiplicadores. Los reactores exponenciales se emplearon despues de construir el PCTR, a pesar de que este cumplio ampliamente sus promesas. El autor proporciona diversos datos tipicos obtenidos con estas dos instalaciones y compara sus papeles respectivos para el estudio de nuevos reactores de potencia, para justificar la

  17. Power Reactor Design at Zero Power; Etudes de Reacteurs de Puissance, au Moyen de Machines de Puissance Zero; Konstruktsiya ehnergeticheskogo reaktora nulevoj moshchnosti; Diseno de Reactores Generadores con Ayuda de Reactores de Potencia Nula

    Energy Technology Data Exchange (ETDEWEB)

    Redman, W. C.; Plumlee, K. E.; Baird, Q. L. [Argonne National Laboratory, Argonne, IL (United States)

    1964-02-15

    de los elementos combustibles de oxido de uranio-oxido de torio en agua pesada, destacando principalmente los datos necesarios para el diseAo de un segundo cuerpo para el reactor experimental de agua hirviente de Argonne; 2. Preparacien de una maqueta de reactor de investigacion de flujo elevado que permitira veriflcar los calculos efectuados durante el estudio, determinar la geometrra optima y evaluar el efecto de la combustion; 3. Determinacion de las distribuciones energeticas y del efecto de inmersion de los elementos combustibles sobre la reactividad en el caso de un reactor de agua hirviente con sobrecalentador incorporado; 4. Diseflo de un cuerpo de reactor reproductor plutonfgeno de neutrones ripidos, refrigerado por sodio y alimentado con {sup 235}U, que constituiri la carga inicial del segundo reactor reproductor experimental (EBR-II) de Argonne; 5. Estudio de las caracterfcticas de un reactor de dos zonas (termica y rapida) que sufren interaccidn. Al discutir estos programas, los autores explican tambien en que factores se basd la eleccion de los experimentos en conjuntos exponenciales y criticos sin envenenamiento en maquetas de potencia nula, asi como de los experimentos in situ, que sirvieron para obtener los datos necesarios. Tambien describen la importancia de los trabajos analiticos complementarios. La memoria presenta ejemplos especfficos para demostrar en que medida se pueden obtener datos sobre el diseno del reactor antes de explotarlo en regimen normal. Entre estos datos se cuenta el margen de paro, el exceso de reactividad necesario para el funcionamiento, los coeficientes de temperatura, la eficacia de las barras de control y de seguridad, la cinetica del reactor, los esquemas de produccion de energia, los requisitos que ha de cumplir la fuente neutronica de puesta en marcha, y la sensibilidad de los instrumentos, los blindajes y la economfa neutronica. El estudio de los experimentos realizados recientemente con reactores de potencia nula

  18. Continuous hyperpolarization with parahydrogen in a membrane reactor

    Science.gov (United States)

    Lehmkuhl, Sören; Wiese, Martin; Schubert, Lukas; Held, Mathias; Küppers, Markus; Wessling, Matthias; Blümich, Bernhard

    2018-06-01

    Hyperpolarization methods entail a high potential to boost the sensitivity of NMR. Even though the "Signal Amplification by Reversible Exchange" (SABRE) approach uses para-enriched hydrogen, p-H2, to repeatedly achieve high polarization levels on target molecules without altering their chemical structure, such studies are often limited to batch experiments in NMR tubes. Alternatively, this work introduces a continuous flow setup including a membrane reactor for the p-H2, supply and consecutive detection in a 1 T NMR spectrometer. Two SABRE substrates pyridine and nicotinamide were hyperpolarized, and more than 1000-fold signal enhancement was found. Our strategy combines low-field NMR spectrometry and a membrane flow reactor. This enables precise control of the experimental conditions such as liquid and gas pressures, and volume flow for ensuring repeatable maximum polarization.

  19. Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

    Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

  20. Nuclear instrumentation systems in prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Vijayakumaran, P.M.; Nagaraj, C.P.; Paramasivan-Pillai, C.; Ramakrishnan, R.; Sivaramakrishna, M.

    2004-01-01

    The nuclear instrumentation systems of the Prototype Fast Breeder Reactor (PFBR) primarily comprise of global Neutron Flux Monitoring, Failed Fuel Detection and Location, Radiation Monitoring and Post-Accident Monitoring. High temperature fission chambers are provided at in-vessel locations for monitoring neutron flux. Failed fuel detection and location is by monitoring the cover gas for fission gases and primary sodium for delayed neutrons. Signals of the core monitoring detectors are used to initiate SCRAM (safety action) to protect the reactor from various postulated initiating events. Radiation levels in all potentially radioactive areas are monitored to act as an early warning system to keep the release of radioactivity to the environment and exposure to personnel well below the permissible limits. Fission Chambers and Gamma Ionisation Chambers are located in the reactor vault concrete for monitoring the neutron flux and gamma radiation levels during and after an accident. (authors)

  1. Management and storage of commercial power reactor wastes

    International Nuclear Information System (INIS)

    1976-01-01

    In May 1976, a technical document, ERDA--76-43, entitled ''Alternatives for Managing Wastes from Reactors and Post-Fission Operations in the LWR Fuel Cycle'' was published by the United States Energy Research and Development Administration. This 1500-page document describes technical alternatives for managing wastes from the commercial light-water-reactor fuel cycle. It does not select preferred waste management technologies or make comparative assessments. This report, ERDA--76-162, is a brief summary of the salient points in the 1500-page document and should provide an appreciation of the present technology and methods for handling the various forms of radioactive waste. In a major expansion of ERDA's waste management program, the U.S. has initiated efforts to identify acceptable geologic formations within the continental U.S. for ultimate disposition of reactor wastes. This technique represents the most advanced alternative presently available for the long-term management of these wastes

  2. The post irradiation examination of three fuel rods from the IFA 429 experiment irradiated in the Halden Reactor

    International Nuclear Information System (INIS)

    Williams, J.

    1979-11-01

    A series of fuel rod irradiation experiments were performed in the Halden Heavy Boiling Water Reactor in Norway. These were designed to provide a range of fuel property data as a function of burn-up. One of these experiments was the IFA-429. This was designed to study the absorption of helium filling gas by the UO 2 fuel pellets, steady state and transient fission gas release and fuel thermal behaviour to high burn-up. This data was to be obtained as a function of fuel density, fuel grain size, initial fuel/cladding gap, average linear heat rating, burn-up and overpower transients. All the fuel is in the form of pressed and sintered UO 2 pellets enriched to 13 weight percent 235 U. All the rods were clad in Zircaloy 4 tube. The details of the experiment are given. The post irradiation examination included: visual examination, neutron radiography, dimensional measurements, gamma scanning, measurement of gases in fuel rods and internal free volume, burn-up analysis, metallographic examination, measurement of retained gas in UO 2 pellets, measurement of bulk density of UO 2 . The results are given and discussed. (U.K.)

  3. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  4. Design criteria for the electrical system in advanced passive reactors. Special features of the AP-600 Reactor

    International Nuclear Information System (INIS)

    Moraleda Lopez, A.

    1997-01-01

    The design of the electrical system of an Passive Advanced Reactor is determined by the concept of passive actuation of safety systems, simplification of process systems and optimisation of equipment performance. The system that results from these criteria is very different to those designed for present plants. The main differences are: No class 1E alternating current systems No emergency diesel generators Fewer safety and non-safety class electricity consumers System for continuous monitoring of battery status Use of electronic speed regulators for reactor feedwater pump motors Outsite battery backup safety power supply Motor-operated valves are the only safety electrical actuators Portable power supply for post 72 hour equipment This paper develops these concepts and applies them to the AP-600 project and describes the electrical system of this type of plant. (Author)

  5. Software for the nuclear reactor dynamics study using time series processing; Software para el estudio de la dinamica de reactores nucleares mediante el procesamiento de series temporales

    Energy Technology Data Exchange (ETDEWEB)

    Valero, Esbel T.; Montesino, Maria E. [Instituto Superior de Ciencia y Tecnologia Nuclear (ISCTN), La Habana (Cuba)

    1997-12-01

    The parametric monitoring in Nuclear Power Plant (NPP) permits the operational surveillance of nuclear reactor. The methods employed in order to process this information such as FFT, autoregressive models and other, have some limitations when those regimens in which appear strongly non-linear behaviors are analyzed. In last years the chaos theory has offered new ways in order to explain complex dynamic behaviors. This paper describes a software (ECASET) that allow, by time series processing from NPP`s acquisition system, to characterize the nuclear reactor dynamic as a complex dynamical system. Here we show using ECASET`s results the possibility of classifying the different regimens appearing in nuclear reactors. The results of several temporal series processing from real systems are introduced. This type of analysis complements the results obtained with traditional methods and can constitute a new tool for monitoring nuclear reactors. (author). 13 refs., 3 figs.

  6. Study on the Post-Fire Safe-Shutdown Analysis for CANDU NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Hwan; Kim, Yun Jung; Park, Mun Hee [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The purpose of this paper is to study a method of the Post-Fire Safe-Shutdown Analysis in order to apply to CANDU NPPs when one group of the Safety Structures, Systems and Components(SCCs) is failed by Fire. The purpose of Fire Protection is prevention, suppression of the fire and mitigation of the effect on the Nuclear Safety. When fire takes place at the Nuclear Power Plants(NPPs), the reactor should achieve and maintain safe shut-down condition and minimize radioactive material release to an environment. The purpose of the Post-Fire SSA process is an evaluation process during a fire at NPPs. At this study, the process was conceptually adopted for control room complex of CANDU NPPs. The Core Damage Frequency of the Reactor will be evaluated more accurately if the SSA is adopted adequately at a fire.

  7. Sedimentación, solubilización, y prefermentación de aguas residuales en un reactor biopelícula

    OpenAIRE

    Cuevas-Rodríguez, Germán; Tejero Monzón, Iñaki

    2003-01-01

    Esta investigación fue realizada con el objetivo de desarrollar un nuevo reactor prefermentador de aguas residuales para incrementar los porcentajes de sedimentación, hidrólisis y prefermentación de la materia orgánica contenida en el agua residual bruta, empleando una sola unidad de pretratamiento y, de esta manera, poder remplazar el decantador primario por este nuevo reactor. El estudio fue realizado en un reactor biopelícula de lecho sumergido fijo, empacado con un medio de soporte BLASF‚...

  8. Manejo enfermero en el síndrome posresección transuretral de próstata (síndrome post-RTUP Syndrome management nurse in post-TURP

    Directory of Open Access Journals (Sweden)

    Manuel Burgos Arguijo

    2011-09-01

    Full Text Available La resección transuretral es una técnica sencilla en cirugía urológica de vías inferiores para tumores de próstata, sin otra patología de vías inferiores, con posibilidad de resección completa, en la cual se utiliza, normalmente, anestesia locorregional, y cuya complicación más importante es el síndrome post-RTUP (posresección transuretral de próstata. Al ser una técnica sencilla y la aplicación de la anestesia ser locorregional, los pacientes intervenidos pasan un corto período de tiempo en la Unidad de Reanimación; es por ello, que el profesional de Enfermería de la Unidad de Cirugía Urológica debe estar familiarizado con una de las complicaciones de la RTUP, como es el síndrome post-RTUP, para ello se llevó a cabo la valoración de las intervenciones realizadas en el Hospital Universitario Severo Ochoa, de Leganés, Madrid, en el año 2008, donde los resultados alcanzaban un 6,49% de pacientes con síndrome post-RTUP, dentro de los referentes bibliográficos establecidos entre el 1% y el 7%; para llegar a la elaboración de un plan de cuidados enfermeros estandarizado en el síndrome post-RTUP, para implementarse en un futuro en nuestra Unidad, teniendo como objetivo general el conocimiento, por parte del personal de enfermería de la misma, de todas las posibles complicaciones que este tipo de intervenciones puede llegar a desarrollar, que sirva además como uno de los protocolos enfermeros del propio centro donde desarrollamos nuestra actividad enfermera, así como que sirva de base para actuaciones futuras en otros centros de nuestro medio.Transurethral resection is a simple technique in lower tract urological surgery for prostate tumors, with no other pathology of the lower airways, with the possibility of complete resection, which is typically used, local anesthesia, and whose most important complication is the syndrome post-TURP (post-TURP. As a simple technique and application of anesthesia to be locoregional

  9. Analysis of a Spanish energy scenario with Generation IV nuclear reactors

    International Nuclear Information System (INIS)

    Ochoa, Raquel; Jimenez, Gonzalo; Perez-Martin, Sara

    2013-01-01

    Highlights: • Spanish energy scenario for the hypothetical deployment of Gen-IV SFR reactors. • Availability of national resources is assessed, considering SFR’s breeding. • An assessment of the impact of transmuting MA on the final repository. • SERPENT code with own pre- and post-processing tools were employed. • The employed SFR core design is based on the specifications of the CP-ESFR. - Abstract: The advantages of fast-spectrum reactors consist not only of an efficient use of fuel through the breeding of fissile material and the use of natural or depleted uranium, but also of the potential reduction of the amount of actinides such as americium and neptunium contained in the irradiated fuel. The first aspect means a guaranteed future nuclear fuel supply. The second fact is key for high-level radioactive waste management, because these elements are the main responsible for the radioactivity of the irradiated fuel in the long term. The present study aims to analyze the hypothetical deployment of a Gen-IV Sodium Fast Reactor (SFR) fleet in Spain. A nuclear fleet of fast reactors would enable a fuel cycle strategy different than the open cycle, currently adopted by most of the countries with nuclear power. A transition from the current Gen-II to Gen-IV fleet is envisaged through an intermediate deployment of Gen-III reactors. Fuel reprocessing from the Gen-II and Gen-III Light Water Reactors (LWR) has been considered. In the so-called advanced fuel cycle, the reprocessed fuel used to produce energy will breed new fissile fuel and transmute minor actinides at the same time. A reference case scenario has been postulated and further sensitivity studies have been performed to analyze the impact of the different parameters on the required reactor fleet. The potential capability of Spain to supply the required fleet for the reference scenario using national resources has been verified. Finally, some consequences on irradiated final fuel inventory are assessed

  10. Biohydrogen production from diary processing wastewater by anaerobic biofilm reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rios-Gonzalez, L.J.; Moreno-Davila, I.M.; Rodriguez-Martinez, J.; Garza-Garcia, Y. [Universidad Autonoma de Coahuila, Saltillo, Coahuila (Mexico)]. E-mail: leopoldo.rios@mail.uadec.mx

    2009-09-15

    to be employed for hydrogen production. [Spanish] Este articulo describe la produccion biologica de hidrogeno a partir de agua residual diaria via fermentacion anaerobica utilizando choque termico pretratado (100 grados centigrados, 30 min.) y procedimientos de tratamiento acido para enriquecer selectivamente el hidrogeno produciendo consorcios mezclados antes de la inoculacion de reactores por lote. El biorreactor empleado para el consorcio de inmovilizacion se opero a temperatura mesofilica (ambiente) (20{+-}3 grados centigrados), bajo condiciones acidofilicas (pH 4.0-4.5), HRT (2h), y un soporte natural para generar hidrogeno produciendo biopelicula de consorcios mezclados: Opuntia imbricata. El reactor se opero inicialmente con sorbitol (5g/L) durante 60 dias de operacion. Las pruebas de lote se llevaron a cabo empleando 20{+-}0.02g de soporte natural con biopelicula. Los experimentos de lote se realizaron para investigar el efecto de la DQO ((2.9-21.1 g-DQO/L), a pH inicial de 7.0, 32{+-}1 grados centigrados. La produccion maxima de hidrogeno se obtuvo a 21.1 g-COD/L. Se efectuaron experimentos del efecto del pH empleando una concentracion de sustrato optima (21.2 g-COD/L), a pH de 4 a 7 y 11.32 (pH de agua residual diaria) y 32{+-}1 grados centigrados. Los resultados de los experimentos indican que el cultivo inicial optimo fue de pH 4.0, pero podemos considerar tambien una produccion estable de hidrogeno a pH 11.32 (pH de agua residual diaria), por lo que se pudo evitar ajustar el pH, y usar agua residual diaria como queda en el proceso de produccion de queso. El pH operacional de 4.0 esta 1.5 unidades por debajo del reportado antes correspondiente al hidrogeno que producen los organismos. La influencia del efecto de la temperatura se realizo usando la concentracion de sustrato optima (21.2 g-COD/L), dos niveles de pH: 4.0 y 11.32, y cuatro diferentes temperaturas: 16{+-}3 grados centigrados (temperatura ambiente), 32{+-}1 grados centigrados, 45{+-}1 grados

  11. Actinide behavior in the integral fast reactor

    International Nuclear Information System (INIS)

    Courtney, J.C.

    1993-05-01

    Goal of this project is to determine the consumption of Np-237, Pu-240, Am-241, and Am-243 in the Integral Fast Reactor (IFR) fuel cycle. These four actinides set the long term waste management criteria for spent nuclear fuel; if it can be demonstrated that they can be efficiently consumed in the IFR, then requirements for nuclear waste repositories can be much less demanding. Irradiations in the Experimental Breeder Reactor II (EBR-II) at Argonne National Laboratory's site near Idaho Falls, Idaho, will be conducted to determine fission and transmutation rates for the four nuclides. The experimental effort involves target package design, fabrication, quality assurance, and irradiation. Post irradiation analyses are required to determine the fission rates and neutron spectra in the EBR-II core

  12. Very-high-temperature gas reactor environmental impacts assessment

    International Nuclear Information System (INIS)

    Baumann, C.D.; Barton, C.J.; Compere, E.L.; Row, T.H.

    1977-08-01

    The operation of a Very High Temperature Reactor (VHTR), a slightly modified General Atomic type High Temperature Gas-Cooled Reactor (HTGR) with 1600 F primary coolant, as a source of process heat for the 1400 0 F steam-methanation reformer step in a hydrogen producing plant (via hydrogasification of coal liquids) was examined. It was found that: (a) from the viewpoint of product contamination by fission and activation products, an Intermediate Heat Exchanger (IHX) is probably not necessary; and (b) long term steam corrosion of the core support posts may require increasing their diameter (a relatively minor design adjustment). However, the hydrogen contaminant in the primary coolant which permeates the reformer may reduce steam corrosion but may produce other problems which have not as yet been resolved. An IHX in parallel with both the reformer and steam generator would solve these problems, but probably at greater cost than that of increasing the size of the core support posts. It is recommended that this corrosion problem be examined in more detail, especially by investigating the performance of current fossil fuel heated reformers in industry. Detailed safety analysis of the VHTR would be required to establish definitely whether the IHX can be eliminated. Water and hydrogen ingress into the reactor system are potential problems which can be alleviated by an IHX. These problems will require analysis, research and development within the program required for development of the VHTR

  13. Design optimization of the Laguna Verde nuclear power station fuel recharge; Optimacion del diseno de recargas de combustible para la Central Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Cortes Campos, Carlos Cristobal [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Montes Tadeo, Jose Luis [Instituto Nacional de Investigaciones Nucleares (ININ), Salazar (Mexico)

    1991-12-31

    It is described, in general terms, the procedure that is followed to accomplish the optimization of the recharge design, and an example is shown where this procedure was applied for the analysis of the type BWR reactor of Unit No. 1 of the Laguna Verde Nuclear Power Station. [Espanol] Se describe en terminos generales el procedimiento que se sigue para realizar la optimacion del diseno de recargas, y se muestra un ejemplo en el que se utilizo dicho procedimiento para el analisis del reactor tipo BWR de la unidad 1, de la Central Laguna Verde (CLV).

  14. Design optimization of the Laguna Verde nuclear power station fuel recharge; Optimacion del diseno de recargas de combustible para la Central Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Cortes Campos, Carlos Cristobal [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Montes Tadeo, Jose Luis [Instituto Nacional de Investigaciones Nucleares (ININ), Salazar (Mexico)

    1992-12-31

    It is described, in general terms, the procedure that is followed to accomplish the optimization of the recharge design, and an example is shown where this procedure was applied for the analysis of the type BWR reactor of Unit No. 1 of the Laguna Verde Nuclear Power Station. [Espanol] Se describe en terminos generales el procedimiento que se sigue para realizar la optimacion del diseno de recargas, y se muestra un ejemplo en el que se utilizo dicho procedimiento para el analisis del reactor tipo BWR de la unidad 1, de la Central Laguna Verde (CLV).

  15. Shadow corrosion evaluation in the Studsvik R2 reactor

    International Nuclear Information System (INIS)

    Sanders, Ch.; Lysell, G.

    2000-01-01

    Post-irradiation examination has shown that increased corrosion occurs when zirconium alloys are in contact with or in proximity to other metallic objects. The observations indicate an influence of irradiation from the adjacent component as the enhanced corrosion occurs as a 'shadow' of the metallic object on the zirconium surface. This phenomenon could ultimately limit the lifetime of certain zirconium alloy components in the reactor. The Studsvik R2 materials test reactor has an In-Core Autoclave (INCA) test facility especially designed for water chemistry and materials research. The INCA facility has been evaluated and found suitable for shadow corrosion studies. The R2 reactor core containing the INCA facility was modeled with the Monte Carlo N-Particle (MCNP) code in order to evaluate the electron deposition in various materials and to develop a hypothesis of the shadow corrosion mechanism. (authors)

  16. Annual report on operation, utilization and technical development of research reactors and hot laboratory

    International Nuclear Information System (INIS)

    1990-09-01

    This report describes the activities of the Department of Research Reactor Operation in fiscal year of 1989. It also presents some technical topics on the reactor operation and utilization in details. The Department is responsible for operation of the research reactors, JRR-2 and JRR-4, and the Hot Laboratory. The research reactor JRR-3 was reconstructed to enhance the performance for utilization. The first criticality was achieved on March 22, 1989, and it subsequently went into operation. In connection with the reactor operation, the various research and development activities in the area of fuel management, water chemistry, radiation monitoring and material irradiation have been made. In the Hot Laboratory, post-irradiation examinations of fuels and materials have been carried out along with the development of related techniques. (author)

  17. Integrated scheme of long-term for spent fuel management of power nuclear reactors; Esquema integrado de largo plazo para la administracion de combustible gastado de reactores nucleares de potencia

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Palacios H, J. C.; Martinez C, E., E-mail: ramon-ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    After of irradiation of the nuclear fuel in the reactor core, is necessary to store it for their cooling in the fuel pools of the reactor. This is the first step in a processes series before the fuel can reach its final destination. Until now there are two options that are most commonly accepted for the end of the nuclear fuel cycle, one is the open nuclear fuel cycle, requiring a deep geological repository for the fuel final disposal. The other option is the fuel reprocessing to extract the plutonium and uranium as valuable materials that remaining in the spent fuel. In this study the alternatives for the final part of the fuel cycle, which involves the recycling of plutonium and the minor actinides in the same reactor that generated them are shown. The results shown that this is possible in a thermal reactor and that there are significant reductions in actinides if they are recycled into reactor fuel. (Author)

  18. Escala de Depressão Pós-natal de Edimburgo para triagem no sistema público de saúde Escala de Depresión Post-natal de Edimburgo para tamizage en el sistema público de salud Edinburgh Postnatal Depression Scale for screening in the public health system

    Directory of Open Access Journals (Sweden)

    Patrícia Figueira

    2009-08-01

    Full Text Available OBJETIVO: Avaliar a utilização da Escala de Depressão Pós-natal de Edimburgo como instrumento de triagem no sistema público de saúde. MÉTODOS: A Escala foi administrada entre o 40º e 90º dia do pós-parto, a 245 mulheres que tiveram parto em uma maternidade privada no município de Belo Horizonte (MG, entre 2005 e 2006. As participantes foram submetidas a uma entrevista psiquiátrica estruturada (Mini-Plus 5.0 utilizada como padrão-ouro para diagnóstico de depressão. Foram calculadas sensibilidade e especificidade da escala e utilizou-se a curva ROC para achar o melhor ponto de corte. Foi utilizado o teste t de Student para comparação das variáveis numéricas e o qui-quadrado para as variáveis categóricas. A confiabilidade foi aferida pelo coeficiente de consistência interna á de Cronbach. RESULTADOS: Foram diagnosticadas 66 mulheres com o quadro depressivo pós-parto (26,9% da amostra. Não houve diferença entre as mulheres com e sem depressão pós-parto em relação à idade, escolaridade, número de partos anteriores e estado civil. Utilizando-se o ponto de corte de 10, a sensibilidade da escala foi 86,4, a especificidade 91,1 e o valor preditivo positivo 0,78. CONCLUSÕES: As propriedades psicométricas da Escala a caracterizam como um bom instrumento de triagem da depressão pós-parto e seu uso disseminado no Sistema Único de Saúde poderia repercutir positivamente com aumento significativo na taxa de reconhecimento, diagnóstico, e tratamento da depressão pós parto.OBJETIVO: Evaluar la utilización de la Escala de Depresión Post-natal de Edimburgo como instrumento de tamizage en el sistema público de salud. MÉTODOS: La Escala fue administrada entre el 40º y 90º día de post-parto, a 245 mujeres que tuvieron parto en una maternidad privada en el municipio de Bello Horizonte (MG, entre 2005 y 2006. Las participantes fueron sometidas a una entrevista psiquiátrica estructurada (Mini-Plus 5.0 utilizada como patr

  19. Reactor core and initially loaded reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

    In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

  20. The Jules Horowitz reactor project, a driver for revival of the research reactor community

    Energy Technology Data Exchange (ETDEWEB)

    Pere, P.; Cavailler, C.; Pascal, C. [AREVA TA, CEA Cadarache - Etablissement d' AREVA TA - Chantier RJH - MOE - BV2 - BP no. 9 - 13115 Saint Paul lez Durance (France); CS 50497 - 1100, rue JR Gauthier de la Lauziere, 13593 Aix en Provence cedex 3 (France)

    2010-07-01

    The first concrete of the nuclear island for the Jules Horowitz Reactor (JHR) was poured at the end of July 2009 and construction is ongoing. The JHR is the largest new platform for irradiation experiments supporting Generation II and III reactors, Generation IV technologies, and radioisotope production. This facility, composed of a unique grouping of workshops, hot cells and hot laboratories together with a first -rate MTR research reactor, will ensure that the process, from preparations for irradiation experiments through post-irradiation non-destructive examination, is completed expediently, efficiently and, of course, safely. In addition to the performance requirements to be met in terms of neutron fluxes on the samples (5x10{sup 14} n.cm{sup -2}/sec{sup -1} E> 1 MeV in core and 3,6x10{sup 14} n.cm{sup -2}/sec{sup -1} E<0.625 eV in the reflector) and the JHR's considerable irradiation capabilities (more than 20 experiments and one-tenth of irradiation area for simultaneous radioisotope production), the JHR is the first MTR to be built since the end of the 1960's, making this an especially challenging project. The presentation will provide an overview of the reactor, hot cells and laboratories and an outline of the key milestones in the project schedule, including initial criticality in early 2014 and radioisotope production in 2015. This will be followed by a description of the project organization set up by the CEA as owner and future operator and AREVA TA as prime contractor and supplier of critical systems, and a discussion of project challenges, especially those dealing with the following items: - accommodation of a broad experimental domain, - involvement by international partners making in-kind contributions to the project, - development of components critical to safety and performance, - the revival of engineering of research reactors and experimental devices involving France's historical players in the field of research reactors, and

  1. POSIBILIDADES DE GOOGLE DRIVE PARA LA DOCENCIA A DISTANCIA Y EN EL AULA

    Directory of Open Access Journals (Sweden)

    EVA MARTÍN RODA

    2015-12-01

    Full Text Available  Resumen: Google Drive es una herramienta de libre acceso en la red, que posibilita la realización de trabajos en línea por parte de los usuarios al almacenarse los documentos en la nube. Dispone de instrumentos como procesador de texto, hojas de cálculo, entre otros. La herramienta es además bastante completa, pues a través de Google Apps permite acceder a numerosas opciones para la labor docente pudiéndose crear tutoriales, grabar vídeos y clases; editar, diseñar y compartir imágenes desde cualquier dispositivo móvil y desde el propio Google drive.  Palabras clave: Google Drive, aprendizaje colaborativo, entorno virtual, educación universitaria. Summary: Google Drive is a free access tool available in the cloud, that allow to the users to do online works and store documents in the cloud. It offers tools such as word processing, spreadsheets, etc... This tool is also quite complete, with several options for teaching as it is possible to createtutorials, or this record videos and classes; edit, design and share pictures from any mobiledevice even from Google Drive. Keywords: Google Drive, collaborative learning, virtual environment, college education. Resumé: Google Drive est un serveur de libre accèsdans la toile, qui facilite auxusagers la réalisation de travaux en ligne grâce aux documents stockés dans le «nuage». Il dispose de plusieur soutilstels qu’unprocesseur de textes, des feuilles de calcul etc… L’outilest en plus assez complet, avec de nombre uses options pour la pratique de l’enseignement car il permet de créer des tutoriels, d’enregistrer des vidéos et des cours, d’éditer, de dessiner et de partager des images depuis n’importe quel dispositif portable ou sur Google Drive directement… Mots clés: Google Drive, apprentissage collaboratif, domaine virtuel, éducation universitaire. 

  2. Study, design and evaluation of nuclear reactor computer control system

    International Nuclear Information System (INIS)

    Menacer, S.

    1988-01-01

    Nuclear reactor control is a complex process that varies with each reactor and there is no universal agreement as to the best type of control system. After the use of conventional systems for a long time, attention turned towards digital techniques in the reactor control system. This interest emerged because of the difficulties faced in the data manipulation, mainly for post-incident analysis. However, it is not sufficient to insert a computer in a system to solve all the data-handling problems and also the insertion of a computer in a real-time system is not without any effect on the overall system. The scope of this thesis is to show the important parameters that have to be taken into account when choosing and evaluate the performances of the selected system

  3. Review of light water reactor safety through the Three Mile Island accident

    International Nuclear Information System (INIS)

    Phung, D.L.

    1984-05-01

    This review of light water reactor safety through the Three Mile Island accident has the purpose of establishing the baseline over which safety achievement post-TMI is assessed, and the need for new reactor designs and business direction is judged. Five major areas of reactor safety pre-TMI are examined: (1) safety philosophy and institutions, (2) reactor design criteria, (3) operational problems, (4) the Rasmussen reactor safety study, and (5) the TMI accident and repercussions. Although nuclear power has made spectacular achievements over the period pre-TMI and although TMI is technically a minor accident, this review concludes that there were basic flaws in the technology and in the manner safety philosophy was conceived and carried out. These flaws included (1) a reactor design that has high core power density, low heat capacity, and low system tolerance to upsets, (2) reactor deployment that had been expedited without extensive operational experience, (3) rules and regulations that had to play catch-up with commercial reactor development, (4) an industry that was fragmented, short-sighted, and tended to rely on the Nuclear Regulatory Commission for safety guidance, (5) information that was not effectively shared, and (6) attention that was inadequate to the human aspects of reactor operation and to public reaction to the specter of a reactor accident, major or minor

  4. New Instruments and Principles for the Dimensional Measurement and Measurement of Spacing of Reactor Components; Nouveaux Instruments et Procedes de Mesure des Dimensions et de l'Espacement des Elements d'un Reacteur; Novye pribory i printsipy izmereniya razmerov i raspolozheniya komponentov reaktora; Nuevos Instrumentos y Principios para Medir las Dimensiones y la Separacion Entre Componentes de Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, P. [Institut Dr. Foerster, Reutlingen, Federal Republic of Germany (Germany)

    1965-09-15

    continu, des dimensions de parties constitutives metalliques de reacteurs et explique diverses methodes de mesure pour les metaux terreux et non terreux (champs magnetiques des courants continus et des courants alternatifs, courants de Foucault). Il decrit des instruments et donne des exemples de mesure telecommandee du diametre, de l'ovalisation, de la distorsion, etc., de diverses pieces; il expose des methodes de mesure de l'espacement des elements de la zone active du reacteur. Le memoire decrit un instrument permettant d'enregistrer le profil de surface et de faire la lecture directe des valeurs de la rugosite (profondeur de rugosite, degre de polissage, direction des irregularites et valeur quadratique moyenne). Il donne des exemples typiques d'emploi de cet instrument pour les pieces d'un reacteur. L'auteur traite en particulier de la possibilite d'utiliser un petit lecteur polyvalent, a l'aide de manipulateurs, dans les zones actives et pour les matieres 'chaudes'. Il discute l'augmentation de la rugosite de surface en fonction de l'accroissement de l'irradiation. (author) [Spanish] Full text: El autor presenta los problemas de medicion del espesor de hojas y de paredes de tubos y recipientes de material austenftico y no ferroso. Se exponen dos metodos para medir el espesor de paredes sin usar elementos en contacto con las mismas: el metodo de las corrientes de Foucault para medir el espesor de hojas y recipientes de material no ferroso y austenftico, empleando bobinas de transicion, y el empleo de corrientes de Foucault para medir espesores de pared en tubos mediante bobinas anulares extensivas. Se describen los instrumentos adecuados y sus aplicaciones. El autor discute ademas la medicion de espesores de pared en componentes no ferrosos para reactores mediante el 'metodo de la esfera magnetica' y explica el principio de este nuevo procedimiento de medicion, se analiza su alcance, sobre todo para mediciones localizadas, y se describe un instrumento utilizado en

  5. Selection of engineering materials and fabrication of liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Patriarca, P.

    1975-01-01

    Information is presented graphically and pictorially concerning the need for nuclear power; basic nuclear concepts including BWR, PWR, HTGR, and LMFBR; the fissioning process; nuclear reactor fuel; fabrication of reactor vessels for LMFBR's; fabrication of intermediate heat exchangers for LMFBR's; piping fabrication for LMFBR's; transition welds; steam generators for LMFBR demonstration plants worldwide; stress corrosion cracking of steam generator materials and weldments; post--test examination of the Alco/BLH sodium-heated steam generator; alternate steam generator designs; and alternate structural materials. (DCC)

  6. Fuel management at the Petten high flux reactor

    International Nuclear Information System (INIS)

    Thijssen, P.J.M.

    1999-01-01

    Several years ago the shipment of spent fuel of the High Flux Reactor (HFR) at Petten has come to a standstill resulting in an ever growing stock of fuel elements that are labelled 'fully burnt up'. Examination of those elements showed that a reasonably number of them have a relatively high 235 U mass left. A reactor physics analysis showed that the use of such elements in the peripheral core zone allows the loading of four instead of five fresh fuel elements in many cycle cores. For the assessment of safety and performance parameters of HFR cores a new calculational tool is being developed. It is based on AEA Technology's Reactor physics code suite Winfrith Improved Multigroup Scheme (WIMS). NRG produced pre- and post-processing facilities to feed input data into WIMS's 2D transport code CACTUS and to extract relevant parameters from the output. The processing facilities can be used for many different types of application. (author)

  7. Determination of the neutron spectrum at different locations in the Argentine RA-1 Reactor; Determinacion del espectro neutronico en distintas posiciones del reactor RA-1

    Energy Technology Data Exchange (ETDEWEB)

    Lerner, A M; Madariaga, M R [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina)

    1999-12-31

    Full text: It is well known that the RA-1 reactor is used to irradiate different types of materials with neutrons. The Radio dosimetry Group (which belongs to the Nuclear Regulatory Authority) uses its fast column for the design, calibration and set up of criticality dosimeters as well as for a quick assessment of the dose to workers in case of an accident. With such purpose, Au(1), Au under Cd and In(2) foils were irradiated to estimate absolute thermal, epithermal and fast neutron fluxes at the irradiation location. The accuracy of this estimation is higher when the response to the present neutron spectrum of the different materials constituting the detectors is better known. This, in turn, requires the previous knowledge of such spectrum (a detailed energy dependence of neutron flux) at the analysed location. In this work a neutronic calculation is presented at the fast irradiation location. The whole calculation was carried out following two different methodologies, and considering a power of 40 kW. The reactor and its surroundings were represented by a simplified one-dimensional model, as a concentric cylindrical set of regions. Figures are drawn representing fast and thermal fluxes (with the cut at 0.4 eV) as a function of the distance to the core centre. The neutron flux (in n/cm{sup 2}sec.eV) as a function of energy is also shown at the fast irradiation location. Values of flux (in n/cm{sup 2}.sec.eV) are also provided as a function of energy in other typical locations, as well as the equivalent integrated flux values (in n/cm{sup 2}.sec). ((1) According to the reaction Au{sup 197}(n,{gamma})Au{sup 198}, having a cross section of {sigma}{sub 0}=98.8b for thermal neutrons. (2) According to the reaction In{sup 115}(n,n`)In{sup 115m}, with a cross section of some 70 mb for neutrons with energies above 1.2MeV). (author) [Espanol] Texto completo: Como se sabe, el reactor RA1 se utiliza para irradiar con neutrones distintos tipos de materiales. El grupo de

  8. Biological oxidation of dissolved methane in effluents from anaerobic reactors using a down-flow hanging sponge reactor.

    Science.gov (United States)

    Hatamoto, Masashi; Yamamoto, Hiroki; Kindaichi, Tomonori; Ozaki, Noriatsu; Ohashi, Akiyoshi

    2010-03-01

    Anaerobic wastewater treatment plants discharge dissolved methane, which is usually not recovered. To prevent emission of methane, which is a greenhouse gas, we utilized an encapsulated down-flow hanging sponge reactor as a post-treatment to biologically oxidize dissolved methane. Within 3 weeks after reactor start-up, methane removal efficiency of up to 95% was achieved with a methane removal rate of 0.8 kg COD m(-3) day(-1) at an HRT of 2 h. After increasing the methane-loading rate, the maximum methane removal rate reached 2.2 kg COD m(-3) day(-1) at an HRT of 0.5 h. On the other hand, only about 10% of influent ammonium was oxidized to nitrate during the first period, but as airflow was increased to 2.5 L day(-1), nitrification efficiency increased to approximately 70%. However, the ammonia oxidation rate then decreased with an increase in the methane-loading rate. These results indicate that methane oxidation occurred preferentially over ammonium oxidation in the reactor. Cloning of the 16S rRNA and pmoA genes as well as phylogenetic and T-RFLP analyses revealed that type I methanotrophs were the dominant methane oxidizers, whereas type II methanotrophs were detected only in minor portion of the reactor. Copyright 2009 Elsevier Ltd. All rights reserved.

  9. Los retos de la gestión financiera frente a la planeación estratégica de las organizaciones y la globalización

    Directory of Open Access Journals (Sweden)

    Germán Guerrero Chaparro

    2003-06-01

    Full Text Available El artículo realiza una reflexión sobre el papel de la gestión financiera en un ambiente cambiante y volátil, donde el criterio de flexibilidad operativa se convierte en uno de los elementos más importantes de creación de valor por parte de las empresas. Ante este escenario, la teoría financiera proporciona el marco conceptual de la transformación de un modelo financiero compatible con la planeación estratégica mediante el uso de las opciones reales. Lo anterior implica un proceso de redireccionamiento en el quehacer de los gerentes financieros, pasando de un procesador de registro de transacciones financieras a un interlocutor para el logro de los objetivos estratégicos de la organización.

  10. EVALUACIÓN DE PROGESTERONA INTRAVAGINAL POST INSEMINACIÓN ARTIFICIAL PARA REDUCIR LA MUERTE EMBRIONARIA EN VACAS

    Directory of Open Access Journals (Sweden)

    Jorge Ignacio Macias

    2014-06-01

    Full Text Available El objetivo de esta investigación fue evaluar el efecto de la administración de un dispositivo intravaginal de progesterona post IATF sobre la reducción de la muerte embrionaria, el cual se aplicó siete días después de la inseminación. Se utilizaron 40 vacas de aptitud cárnica de cruza mestiza Senangus x Brahman de dos-cuatro partos. Se realizó ecografía transrectal (Aquilla Vet 7.5 MHZ previo a la sincronización. Se efectuó el inicio de ésta con el dispositivo CIDR 1.38 g progesterona. Después de siete días de la inseminación se colocó el dispositivo usado a 20 animales tratamiento y se lo mantuvo durante 10 días. Se realizó tomas de muestras de sangre para medir los niveles de P4. Estas fueron tomadas a cinco animales del grupo tratamiento en el día que se reinsertó el dispositivo, 5 días después y al retiro. Las muestras revelaron niveles altos de progesterona que oscilaron entre 7.87 hasta 19.4 ng/mL. Se diagnosticó gestación a los 30 y 60 días. Se realizó un análisis costo beneficio al tratamiento. En los resultados se encontró una relación entre los tratamientos y la preñez a los 30 días (p=0.053; la relación del costo beneficio se manifestó económicamente viable con la aplicación de progesterona, ya que por cada dólar invertido se genera como ganancia 23 ctvs de dólar. Se concluye que la aplicación de progesterona reduce la muerte embrionaria hasta los 30 días post inseminación.

  11. AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Phil [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ploger, Scott A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-12-01

    The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance, and fission product transport (INL 2015). A series of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously.

  12. AGR-1 Compact 1-3-1 Post-Irradiation Examination Results

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-12-01

    The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport (INL 2015). A series of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously (Grover, Petti, and Maki 2010, Maki 2009).

  13. Imperial College Reactor Centre annual report. 1983

    International Nuclear Information System (INIS)

    1984-01-01

    It is reported that the reactor operated reliably during the year with less than half a day of operating time lost by faults or failures. Brief accounts of the 34 research projects at the Centre are given, and a list of teaching experiments or visits is included. These include undergraduate and post-graduate teaching. Commercial requests for irradiations and neutron activation analysis are reported as increasing. (U.K.)

  14. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  15. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

    A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

  16. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  17. Group cross-sections for fast reactors; Sections efficaces de groupes pour les reacteurs a neutrons rapides; Gruppovye secheniya reaktorov na bystrykh nejtronakh; Secciones eficaces de grupos para reactores rapidos

    Energy Technology Data Exchange (ETDEWEB)

    Zweifel, P P [University of Michigan, Ann Arbor, MI (United States); Ball, G L [Atomic Power Development Associates, Inc., Detroit, MI (United States)

    1962-03-15

    , comme c'est souvent le cas, la section efficace de groupe en terme d'integrales de resonance efficace, mais qu'il faut modifier cette definition suivant le type de schema multigroupe utilise. (author) [Spanish] La memoria discute en terminos generales las ecuaciones de difusion de grupos multiples y la forma correcta de las secciones eficaces correspondientes . En particular, demuestra que la seccion eficaz media de transporte puede expresarse con bastante precision en terminos de un promedio de recorridos libres medios. Esta magnitud es dificil de calcular porque no se puede expresar en funcion de promedios elementales ; sin embargo, se demuestran varias desigualdades que simplifican el procedimiento de determinacion de promedios. La memoria discute otros tres aspectos de las secciones eficaces de grupos que con frecuencia se ignoran, pero que pueden ser importantes al estudiar detalladamente un diseno. a) El empleo de los mismos valores medios correspondientes a las secciones eficaces de grupos para todos los reactores rapidos no se justifica si los espectros de los diferentes reactores no son similares y si las secciones eficaces varian rapidamente dentro del grupo, como ocurre a menudo. Los autores describen un metodo de iteracion, que permite obtener valores medios correctos y determinar en que medida los efectos espectrales ejercen influencia sobre los calculos de reactores. b) En los calculos de transporte (metodo S{sub n} por ejemplo), los promedios deben evaluarse en funcion del angulo y de la energia. Como el flujo no es separable en una parte angulo y en una parte energetica, es necesario proceder con sumo cuidado para evitar errores. La ecuacion S{sub n} se estudia sobre la base de un modelo sencillo, y de este estudio se deduce un criterio que puede ser de utilidad al determinar la importancia de la no-separabilida d angular en los calculos de reactores. c) Basandose en los argumentos de conservacion neutronica, se deriva una relacion de compatibilidad

  18. The Brazilian Multipurpose Reactor (RMB) Project

    Energy Technology Data Exchange (ETDEWEB)

    Perrotta, Jose Augusto [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2012-07-01

    Full text: The Plan of Action on Science, Technology and Innovation (PACT 2007-2010) of the Ministry of Science Technology and Innovation (MCTI), aligned to the governmental strategies for the Brazilian Nuclear Program, established as a goal the study and definition of the Brazilian Multipurpose Reactor (RMB). The RMB research reactor is designed to perform three main functions: radioisotope production for medicine, industry, agriculture and environmental applications; fuel and material irradiation testing in support to the Brazilian nuclear energy program; and to provide neutron beams for scientific and applied research. The main project facilities are: nuclear pool type reactor with a flux level compatible to the multipurpose uses; hot cells laboratory for Mo-99 and I-131 processing; hot cells laboratory for radioisotope processing; hot cells laboratory for irradiated material post irradiation analysis; neutron beams laboratory building with scientific equipment and instrumentation for researching; radiochemistry laboratory; radioactive waste treatment facility; support laboratories for operation and researching; and buildings for researchers and operators. This speech presents the RMB project status, giving some technical and management details on its development and its future perspectives for new jobs in research activities for the Brazilian technical and scientific community. (author)

  19. The Brazilian Multipurpose Reactor (RMB) Project

    International Nuclear Information System (INIS)

    Perrotta, Jose Augusto

    2012-01-01

    Full text: The Plan of Action on Science, Technology and Innovation (PACT 2007-2010) of the Ministry of Science Technology and Innovation (MCTI), aligned to the governmental strategies for the Brazilian Nuclear Program, established as a goal the study and definition of the Brazilian Multipurpose Reactor (RMB). The RMB research reactor is designed to perform three main functions: radioisotope production for medicine, industry, agriculture and environmental applications; fuel and material irradiation testing in support to the Brazilian nuclear energy program; and to provide neutron beams for scientific and applied research. The main project facilities are: nuclear pool type reactor with a flux level compatible to the multipurpose uses; hot cells laboratory for Mo-99 and I-131 processing; hot cells laboratory for radioisotope processing; hot cells laboratory for irradiated material post irradiation analysis; neutron beams laboratory building with scientific equipment and instrumentation for researching; radiochemistry laboratory; radioactive waste treatment facility; support laboratories for operation and researching; and buildings for researchers and operators. This speech presents the RMB project status, giving some technical and management details on its development and its future perspectives for new jobs in research activities for the Brazilian technical and scientific community. (author)

  20. Post accidental small breaks analysis

    International Nuclear Information System (INIS)

    Depond, G.; Gandrille, J.

    1980-04-01

    EDF ordered to FRAMATOME by 1977 to complete post accidental long term studies on 'First Contrat-Programme' reactors, in order to demonstrate the safety criteria long term compliance, to get information on NSSS behaviour and to improve the post accidental procedures. Convenient analytical models were needed and EDF and FRAMATOME respectively developped the AXEL and FRARELAP codes. The main results of these studies is that for the smallest breaks, it is possible to manually undertake cooling and pressure reducing actions by dumping the steam generators secondary side in order to meet the RHR operating specifications and perform long term cooling through this system. A specific small breaks procedure was written on this basis. The EDF and FRAMATOME codes are continuously improved; the results of a French set of separate effects experiments will be incorporated as well as integral system verification

  1. Balanceo de un sistema de cosecha mecanizado utilizando simulación de eventos discretos.

    Directory of Open Access Journals (Sweden)

    Pablo Aracena

    2010-08-01

    Full Text Available Se analizó un sistema de cosecha mecanizado, operando en una faena a tala rasa de pino radiata y conformado por un feller buncher, un skidder con garra, un procesador y un trineumático. Un Modelo de Simulación de Eventos Discretos (MSED fue desarrollado con el propósito de balancear el sistema. El proceso mecanizado de madera fue el limitante del sistema de acuerdo con los resultados del estudio de tiempos, entonces este proceso fue apoyado agregando a 3 operadores de motosierra para alcanzar una producción de 60,82 m3/Hora-Máquina (HM en la configuración propuesta. Simulado el madereo se determinó que para balancear la producción obtenida en proceso, el skidder con garra debía operar a una distancia promedio de madereo (DPM de 80 m, realizando una detención para completar una carga de cuatro fustes. Finalmente, el feller buncher puede usar hasta 1,6 minutos adicionales para formar las pilas que permitan lograr la producción esperada para el sistema. De acuerdo con la experimentación de simulación se podría lograr una producción del sistema de 60,39 m3/HM, si todas las modificaciones propuestas en volteo, apilado, madereo y proceso son realizadas.

  2. Development of core thermal-hydraulics module for intelligent reactor design system (IRDS)

    International Nuclear Information System (INIS)

    Kugo, Teruhiko; Nakagawa, Masayuki; Fujii, Sadao.

    1994-08-01

    We have developed an innovative reactor core thermal-hydraulics module where a designer can easily and efficiently evaluate his design concept of a new type reactor in the thermal-hydraulics field. The main purpose of this module is to decide a feasible range of basic design parameters of a reactor core in a conceptual design stage of a new type reactor. The module is to be implemented in Intelligent Reactor Design System (IRDS). The module has the following characteristics; 1) to deal with several reactor types, 2) four thermal hydraulics and fuel behavior analysis codes are installed to treat different type of reactors and design detail, 3) to follow flexibly modification of a reactor concept, 4) to provide analysis results in an understandable way so that a designer can easily evaluate feasibility of his concept, and so on. The module runs on an engineering workstation (EWS) and has a user-friendly man-machine interface on a pre- and post-processing. And it is equipped with a function to search a feasible range called as Design Window, for two design parameters by artificial intelligence (AI) technique and knowledge engineering. In this report, structure, guidance for users of an usage of the module and instruction of input data for analysis modules are presented. (author)

  3. Description of the blowdown test facility COG program on in-reactor fission product release, transport, and deposition under severe accident conditions

    International Nuclear Information System (INIS)

    Fehrenbach, P.J.; Wood, J.C.

    1987-06-01

    Loss-of-coolant accidents with additional impairment of emergency cooling would probably result in high fuel temperatures leading to severe fuel damage (SFD) and significant fission product activity would then be transported along the PHTS to the break where a fraction of it would be released and transport under such conditions, there are many interacting and sometimes competing phenomena to consider. Laboratory simulations are being used to provide data on these individual phenomena, such as UO 2 oxidation and Zr-UO 2 interaction, from which mathematical models can be constructed. These are then combined into computer codes to include the interaction effects and assess the overall releases. In addition, in-reactor tests are the only source of data on release and transport of short-lived fission product nuclides, which are important in the consequence analysis of CANDU reactor accidents. Post-test decontamination of an in-reactor test facility also provides a unique opportunity to demonstrate techniques and obtain decontamination data relevant to post-accident rehabilitation of CANDU power reactors. Specialized facilities are required for in-reactor testing because of the extensive release of radioactive fission products and the high temperatures involved (up to 2500 degrees Celsius). To meet this need for the Canadian program, the Blowdown Test Facility (BTF) has been built in the NRU reactor at Chalk River. Between completion of construction in mid-1987 and the first Zircaloy-sheathed fuel test in fiscal year 1987/88, several commissioning tests are being performed. Similarly, extensive development work has been completed to permit application of instrumentation to irradiated fuel elements, and in support of post-test fuel assembly examination. A program of decontamination studies has also been developed to generate information relevant to post-accident decontamination of power reactors. The BTF shared cost test program funded by the COG High Temperature

  4. Empleo de una sonda infrarroja in situ para monitorear reacciones de esterificación

    Directory of Open Access Journals (Sweden)

    Francisco José Sánchez Castellanos

    2006-01-01

    Full Text Available Se empleó un reactor batch (por lotes, dotado de tres detectores: pH, Sonda IR y operación en continuo, de tal forma que puede operarse como un reactor CSTR. En la medida en que la esterificación procede, decrecen las bandas correspondientes al grupo -COOH del ácido carboxIlico y la del grupo C-OH del alcohol, presentándose al mismo tiempo incremento en la banda del grupo -COOR del ester que se está formando. El progreso de la reacción se puede seguir por el registro continuo de los espectro IR. La banda correspondiente a H-O-H del agua no se puede seguir ya que se requiere de un ambiente absolutamente anhidro para hacerlo. De otro lado, por aparte pueden prepararse soluciones patrones para poder cuantificar la intensidad de los picos en el espectro IR, segün la composición del componente en la mezcla. Sin embargo, cuando se presentan cambios de fase en la mezcla reactiva, este metodo no puede emplearse para seguir el curso de una reacción, ya que se presenta una variación muy aleatoria en la senal de intensidad de los picos.

  5. Fast Reactor Safety Research Program. Quarterly report, January--March 1976

    International Nuclear Information System (INIS)

    1976-07-01

    Progress is summarized in the following study areas: (1) prompt burst excursion, (2) post-accident heat removal (PAHR) debris bed, (3) fuel motion detection, (4) PAHR molten pool behavior, (5) equation-of-state high-temperature fuel vapor data, and (6) fuel motion detection equipment for the upgraded Annular Core Pulsed Reactor

  6. First fuel re-load of Angra-1 reactor - Inspection and hearing plan

    International Nuclear Information System (INIS)

    Pollis, W.; Alvarenga, M.A.B.; Meldonian, N.L.; Paiva, R.L.C. de; Pollis, R.

    1985-01-01

    The plan of inspection and hearing of the first fuel reload of Angra-1 nuclear reactor is detailed. It consists in five steps: receiving and storage of the fuel; reload preparation; activities during; post-reload activities, and preliminary activities. (M.I.)

  7. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

    1987-01-01

    Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

  8. Estudio comparativo de dos métodos de extracción del nivel de nitrógeno ureico post hemodiálisis para el cálculo del kt/v

    Directory of Open Access Journals (Sweden)

    José Luis Cobo Sánchez

    2013-12-01

    Full Text Available Objetivo: Evaluar si existen diferencias en el cálculo del Kt/V entre 2 métodos de extracción del BUN post HD. Metodología: Estudio prospectivo comparativo en una cohorte de pacientes en hemodiálisis crónica. Se cuantificó el Kt/V durante dos semanas consecutivas mediante la fórmula de Daugirdas de segunda generación. La extracción de la muestra de sangre para el BUN post hemodiálisis se realizó mediante 2 métodos: disminución del flujo de bomba de sangre a 50ml/min durante 2 minutos, inmediatamente antes de finalizar la sesión (método A; y a los 10 minutos tras finalizar la sesión de hemodiálisis (método B. Resultados: 47 pacientes estudiados: 66% hombres, edad media de 66±13 años, 51% FAVI, 59,5% hemodiafiltración on-line. La media del Kt/V durante las dos semanas del estudio para el método A fue de 1,51 y para el método B de 1,41 (p<0,001. Existieron diferencias estadísticamente significativas según la técnica de hemodiálisis y el acceso vascular entre ambos métodos de extracción. El coeficiente de correlación de Pearson sin embargo, mostró una correlación casi lineal entre las medias del Kt/V de ambos métodos (r=0.954. La diferencia entre ambos métodos fue del 10%. Conclusiones: Existen diferencias significativas en el cálculo del Kt/V entre la extracción del BUN post hemodiálisis por el método A y B, siendo mayor la dosis obtenida por el primer método. Sin embargo, ambas mediciones se correlacionan bien, teniendo en cuenta que con la extracción a los 10 minutos se produce una disminución del 10% sobre el otro método.

  9. Impact of Pre-Initiators on PSA in Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ochirbat, Chimedtseren [KAIST, Daejeon (Korea, Republic of); Kim, Sok Chul [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-10-15

    Most of nuclear power plants had already conducted PSA work to examine their plant safety for identifying vulnerability and preparing the mitigating strategies for severe accident. However, the PSA for research reactor has been conducted limitedly comparing with nuclear power plants due to lack of awareness and resources. Most of PSA results demonstrated that human failure events (HFEs) take a major role of risk contributor in terms of core damage frequency. HFEs are categorized as the following three types: pre-initiating event interaction (e.g., maintenance of errors, testing errors, calibration errors), initiating event related interactions (e.g., human error causing loss of power, human error causing system trip), and post-initiating event (e.g., all action actuating manual safety system backup of an automatic system). Lack of resources and utilization of research reactor calls a vicious circle in terms of safety degradation. The safety degradation poses the vulnerability of human failure during research reactor utilization process. Typically, evaluation of pre-initiators related to test and maintenance are not taking into account in PSA for research reactors. This paper aims to investigate the impact of pre-initiating events related to test and maintenance activities on PSA results in terms of core damage frequency for a research reactor.

  10. Impact of Pre-Initiators on PSA in Research Reactor

    International Nuclear Information System (INIS)

    Ochirbat, Chimedtseren; Kim, Sok Chul

    2014-01-01

    Most of nuclear power plants had already conducted PSA work to examine their plant safety for identifying vulnerability and preparing the mitigating strategies for severe accident. However, the PSA for research reactor has been conducted limitedly comparing with nuclear power plants due to lack of awareness and resources. Most of PSA results demonstrated that human failure events (HFEs) take a major role of risk contributor in terms of core damage frequency. HFEs are categorized as the following three types: pre-initiating event interaction (e.g., maintenance of errors, testing errors, calibration errors), initiating event related interactions (e.g., human error causing loss of power, human error causing system trip), and post-initiating event (e.g., all action actuating manual safety system backup of an automatic system). Lack of resources and utilization of research reactor calls a vicious circle in terms of safety degradation. The safety degradation poses the vulnerability of human failure during research reactor utilization process. Typically, evaluation of pre-initiators related to test and maintenance are not taking into account in PSA for research reactors. This paper aims to investigate the impact of pre-initiating events related to test and maintenance activities on PSA results in terms of core damage frequency for a research reactor

  11. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    1980-01-01

    The performance test on the reactor power increase to 75 MW was started on July 3, and the target of 75 MW was reached on July 16, in the experimental fast reactor Joyo. The tests on the heat transport characteristics, power coefficient, the response to the change of outlet temperature, the loss of external power supply and so on were carried out, and the performance test was finished on August 23, except the test of 75 MW continuous operation. The annual inspection of the systems is being carried out in parallel with the regular inspection. The design to prepare for the manufacture of the prototype fast reactor Monju is being prepared. The analysis of decay heat removal is being carried out, and various calculation codes were developed. The technological survey on overseas LMFBRs is being made. The conceptual design of the demonstration reactor is being prepared. The research and development of reactor physics, structural components for Joyo and Monju, instrumentation and control, sodium technology, fuel materials, structural materials, safety problems and steam generators are reported. The tests on the transient boiling of sodium, fuel failure propagation, heat transfer between molten materials, post-accident decay heat removal and so on have been carried out. (Kako, I.)

  12. Ageing Management for Research Reactors. Specific Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

    This Safety Guide was developed under the IAEA programme for safety standards for research reactors, which covers all the important areas of research reactor safety. It supplements and elaborates upon the safety requirements for ageing management of research reactors that are established in paras 6.68-6.70 and 7.109 of the IAEA Safety Requirements publication, Safety of Research Reactors. The safety of a research reactor requires that provisions be made in its design to facilitate ageing management. Throughout the lifetime of a research reactor, including its decommissioning, ageing management of its structures, systems and components (SSCs) important to safety is required, to ensure continued adequacy of the safety level, reliable operation of the reactor, and compliance with the operational limits and conditions. Managing the safety aspects of research reactor ageing requires implementation of an effective programme for the monitoring, prediction, and timely detection and mitigation of degradation of SSCs important to safety, and for maintaining their integrity and functional capability throughout their service lives. Ageing management is defined as engineering, operation, and maintenance strategy and actions to control within acceptable limits the ageing degradation of SSCs. Ageing management includes activities such as repair, refurbishment and replacement of SSCs, which are similar to other activities carried out at a research reactor in maintenance and testing or when a modification project takes place. However, it is important to recognize that effective management of ageing requires the use of a methodology that will detect and evaluate ageing degradation as a consequence of the service conditions, and involves the application of countermeasures for prevention and mitigation of ageing degradation. The objective of this Safety Guide is to provide recommendations on managing ageing of SSCs important to safety at research reactors on the basis of international

  13. Ageing Management for Research Reactors. Specific Safety Guide

    International Nuclear Information System (INIS)

    2010-01-01

    This Safety Guide was developed under the IAEA programme for safety standards for research reactors, which covers all the important areas of research reactor safety. It supplements and elaborates upon the safety requirements for ageing management of research reactors that are established in paras 6.68-6.70 and 7.109 of the IAEA Safety Requirements publication, Safety of Research Reactors. The safety of a research reactor requires that provisions be made in its design to facilitate ageing management. Throughout the lifetime of a research reactor, including its decommissioning, ageing management of its structures, systems and components (SSCs) important to safety is required, to ensure continued adequacy of the safety level, reliable operation of the reactor, and compliance with the operational limits and conditions. Managing the safety aspects of research reactor ageing requires implementation of an effective programme for the monitoring, prediction, and timely detection and mitigation of degradation of SSCs important to safety, and for maintaining their integrity and functional capability throughout their service lives. Ageing management is defined as engineering, operation, and maintenance strategy and actions to control within acceptable limits the ageing degradation of SSCs. Ageing management includes activities such as repair, refurbishment and replacement of SSCs, which are similar to other activities carried out at a research reactor in maintenance and testing or when a modification project takes place. However, it is important to recognize that effective management of ageing requires the use of a methodology that will detect and evaluate ageing degradation as a consequence of the service conditions, and involves the application of countermeasures for prevention and mitigation of ageing degradation. The objective of this Safety Guide is to provide recommendations on managing ageing of SSCs important to safety at research reactors on the basis of international

  14. Advanced Test Reactor National Scientific User Facility Partnerships

    International Nuclear Information System (INIS)

    Marshall, Frances M.; Allen, Todd R.; Benson, Jeff B.; Cole, James I.; Thelen, Mary Catherine

    2012-01-01

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin

  15. 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)

  16. Efecto de dos metales pesados, cadmio y níquel, sobre la eficiencia de remoción de carga orgánica de un reactor UASB a escala de laboratorio

    Directory of Open Access Journals (Sweden)

    Luis Eduardo Forero

    2004-01-01

    Full Text Available Se realizaron ensayos en tres reactores UASB de tres litros cada uno, a un tiempo de retención hidráulico (TRH de cuatro horas y carga orgánica volumétrica de 4,8 g/L/d. Después de la fase inicial de arranque, tiempo de 4.000 horas para los tres reactores, se procedió a afectarlos de la siguiente forma: el primer reactor fue alimentado con 5 mg/L de cloruro de cadmio en forma continua, el segundo reactor fue alimentado con 10 mg/L de cloruro de níquel en forma continua también, mientras que el tercer reactor no se afectó con sustancia alguna y sirvió como control. La eficiencia de remoción de demanda química de oxígeno (DQO del primer reactor cambió del 60% de la fase de arranque (fase 1 al 18% en la fase afectada con cadmio (fase dos; la eficiencia de remoción de DQO en el reactor dos pasó del 60 al 24% y a su vez para el reactor tres control no hubo cambio significativo en dicha eficiencia. A su vez el reactor uno acumuló el cadmio en el lodo, mientras que el reactor dos no hizo lo propio con el níquel.

  17. Steam leak detection in advance reactors via acoustics method

    International Nuclear Information System (INIS)

    Singh, Raj Kumar; Rao, A. Rama

    2011-01-01

    Highlights: → Steam leak detection system is developed to detect any leak inside the reactor vault. → The technique uses leak noise frequency spectrum for leak detection. → Testing of system and method to locate the leak is also developed and discussed in present paper. - Abstract: Prediction of LOCA (loss of coolant activity) plays very important role in safety of nuclear reactor. Coolant is responsible for heat transfer from fuel bundles. Loss of coolant is an accidental situation which requires immediate shut down of reactor. Fall in system pressure during LOCA is the trip parameter used for initiating automatic reactor shut down. However, in primary heat transport system operating in two phase regimes, detection of small break LOCA is not simple. Due to very slow leak rates, time for the fall of pressure is significantly slow. From reactor safety point of view, it is extremely important to find reliable and effective alternative for detecting slow pressure drop in case of small break LOCA. One such technique is the acoustic signal caused by LOCA in small breaks. In boiling water reactors whose primary heat transport is to be driven by natural circulation, small break LOCA detection is important. For prompt action on post small break LOCA, steam leak detection system is developed to detect any leak inside the reactor vault. The detection technique is reliable and plays a very important role in ensuring safety of the reactor. Methodology developed for steam leak detection is discussed in present paper. The methods to locate the leak is also developed and discussed in present paper which is based on analysis of the signal.

  18. Post 9-11 Security Issues for Non-Power Reactor Facilities

    International Nuclear Information System (INIS)

    Zaffuts, P. J.

    2003-01-01

    This paper addresses the legal and practical issues arising out of the design and implementation of a security-enhancement program for non power reactor nuclear facilities. The security enhancements discussed are derived from the commercial nuclear power industry's approach to security. The nuclear power industry's long and successful experience with protecting highly sensitive assets provides a wealth of information and lessons that should be examined by other industries contemplating security improvements, including, but not limited to facilities using or disposing of nuclear materials. This paper describes the nuclear industry's approach to security, the advantages and disadvantages of its constituent elements, and the legal issues that facilities will need to address when adopting some or all of these elements in the absence of statutory or regulatory requirements to do so

  19. Post 9-11 Security Issues for Non-Power Reactor Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Zaffuts, P. J.

    2003-02-25

    This paper addresses the legal and practical issues arising out of the design and implementation of a security-enhancement program for non power reactor nuclear facilities. The security enhancements discussed are derived from the commercial nuclear power industry's approach to security. The nuclear power industry's long and successful experience with protecting highly sensitive assets provides a wealth of information and lessons that should be examined by other industries contemplating security improvements, including, but not limited to facilities using or disposing of nuclear materials. This paper describes the nuclear industry's approach to security, the advantages and disadvantages of its constituent elements, and the legal issues that facilities will need to address when adopting some or all of these elements in the absence of statutory or regulatory requirements to do so.

  20. Post-irradiation studies of test plates for low enriched fuel elements for research reactors

    International Nuclear Information System (INIS)

    Groos, E.; Buecker, H.J.; Derz, H.; Schroeder, R.

    1988-07-01

    In developing new fuels for research reactor elements that allow the use of low enriched uranium (LEU) 3 Si 2 , U 3 Si 1.5 , U 3 Si 1.3 and U 3 Si. Even up to high burnup rates (80% fifa) U 3 Si 2 was proved to be a reliable fuel that according to the test results achieved to date complies with all necessary requirements above all with respect to dimensional stability. U 3 Si showed significant changes of the fuel microstructure associated with considerably higher fuel swelling, that will probably exclude its use in research reactor operation. The irradiation of U 3 Si 1.3 and U 3 Si 1.5 plates had to be terminated untimely. Up to a burnup of 40% fifa these plates behaved quite well. An extrapolation to higher burnup rates, however only seems to be possible with reservations. (orig./HP) [de

  1. New digital control system for the operation of the Colombian research reactor IAN-R1; Nuevo sistema de control digital para la operacion del reactor de investigacion Colombiano IAN-R1

    Energy Technology Data Exchange (ETDEWEB)

    Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J., E-mail: lina.celis@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

  2. Simulación CFD de la transferencia de calor en un reactor de hidrotratamiento de aceites vegetales de segunda generación

    OpenAIRE

    Mendoza Sépulveda, César Camilo

    2013-01-01

    Resumen: Se desarrolló un modelo CFD que permite representar la transferencia de calor en un reactor de hidrotratamiento de aceites vegetales. Este modelo permitió evaluar la transferencia de calor para distintas configuraciones del reactor. En el proceso de hidrotratamiento de aceites vegetales se transforma el aceite en un líquido con cero contenido de azufre y excelentes propiedades como combustible diesel. El proceso se basa en la adición de hidrógeno a alta presión en un reactor de lecho...

  3. Tight aspect ratio tokamak power reactor with superconducting TF coils

    International Nuclear Information System (INIS)

    Nishio, S.; Tobita, K.; Konishi, S.; Ando, T.; Hiroki, S.; Kuroda, T.; Yamauchi, M.; Azumi, M.; Nagata, M.

    2003-01-01

    Tight aspect ratio tokamak power reactor with super-conducting toroidal field (TF) coils has been proposed. A center solenoid coil system and an inboard blanket were discarded. The key point was how to find the engineering design solution of the TF coil system with the high field and high current density. The coil system with the center post radius of less than 1 m can generate the maximum field of ∼ 20 T. This coil system causes a compact reactor concept, where the plasma major and minor radii of 3.75 m and 1.9 m, respectively and the fusion power of 1.8 GW. (author)

  4. The role of post accident chemistry data in nuclear safety

    International Nuclear Information System (INIS)

    Bradshaw, R.W.; Caruthers, G.F.

    1982-01-01

    The NRC instituted the NUREG-0737 requirements as implementation of the Post-TMI Action Plan in October, 1980. Among these requirements was the capability to obtain chemistry samples of the reactor coolant and containment building atmosphere under post accident conditions. The quantitative criteria were, in general, beyond the capabilities of existing plant systems. As a consequence the nuclear industry expended substantial efforts to design and install the post-accident sampling systems necessary to comply with these criteria. With such efforts essentially complete, the task remains to establish the role that data provided by these systems would play in mitigating the consequences of a nuclear plant accident. This role definition must include a characterization of the timing and priority for the post accident chemistry data. This paper defines that role using the Safety Level and Safety Function concepts as a matrix

  5. The electronuclear cycle: from fission to new reactor systems

    International Nuclear Information System (INIS)

    Belier, G.; Cugnon, J.; Lapoux, V.; Liatard, E.; Porquet, Marie-Genevieve; Rudolf, G.

    2006-09-01

    The Joliot Curie School trains each year, and since 1981, PhD students, post-Doctorates and researchers on scientific breakthroughs performed in a topic related to nuclear physics, in a broad range. These proceedings brings together the 11 lectures given at the 2006 session of Joliot Curie School on the topic of the electronuclear cycle: - Fission: from phenomenology to theory (Berger, J.F.); - Physics of nuclear reactors (Baeten, P.); - Data modeling and evaluation (Bauge, E.; Hilaire, S.); - Measurement of cross sections of interest for minor actinides incineration (Jurado, B.); - Spallation data and modelling for hybrid reactors (Boudard, A.); - Nuclear wastes: overview (Billard, I.); - Long living nuclear wastes transmutation processes and feasibility (Varaine, F.); - Hybrid reactors: recent advances for a demonstrator (Billebaud, A.); - Systems of the future and strategy (David, S.); - Non-nuclear energies (Nifenecker, H.); - Fundamental physics with ultracold neutrons (Protasov, K). The last section is a compilation of abstracts of presentations given by Young searchers' (Young searchers' seminars)

  6. Effect of the empty fraction in a solar reactor of fluidized bed; Efecto de la fraccion vacia en un reactor solar de lecho fluidizado

    Energy Technology Data Exchange (ETDEWEB)

    Torres, Alejandro; Romero-Paredes, Hernando; Vazquez, Alejandro; Torijano, Eugenio; Ambriz, Juan J [Universidad Autonoma Metropilitana-Iztapalapa, Mexico, D.F. (Mexico)

    2000-07-01

    de lecho fluidizado que a la vez sirve como receptor solar de un sistema de almacenamiento termoquimico de la energia solar. Los complejos fenomenos que se tienen en estos reactores hacen dificil su dimensionamiento para su aplicacion solar. Por ello, modelar y simular su comportamiento sin y con reaccion quimica ayuda a paliar este inconveniente. Uno de los fenomenos presentes es el cambio de la fraccion vacia en la cual concentramos nuestra atencion. En este trabajo se propone una alternativa en el modelado de estos sistemas considerando las fluctuaciones locales de la fraccion vacia o porosidad {epsilon}(x, y) en el lecho. Para esto se propone una distribucion probabilistica uniforme para todos los nodos de la malla (x, y) del lecho donde se asocian valores de porosidad locales para cada nodo de la malla mediante un generador aleatorio donde {epsilon}(x, y){epsilon} [0.1]. La fraccion nueva juega un papel muy importante debido a que la penetracion de la radiacion solar en estos sistemas de cuerpos opacos depende directamente de la distribucion de espacios vacios en la trayectoria de la radiacion incidente que afecta su comportamiento termico y cinetico. De los resultados se puede comprobar la caracteristica de no isotermicidad del reactor lo que conlleva, una vez alcanzada la temperatura de reaccion, a un perfil de concentraciones disperso. La fraccion vacia es un parametro que influye grandemente en estos perfiles y que incrementando el numero de fluidizacion es como se logra disminuir ese tiempo. En conclusion, se resalta la importancia que juega la fraccion vacia en la evolucion tanto de los perfiles de temperatura como de concentracion. El comportamiento del lecho en la simulacion se hace mas precisa de acuerdo con los resultados experimentales previamente obtenidos.

  7. Post irradiation test report of irradiated DUPIC simulated fuel

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Jung, I. H.; Moon, J. S. and others

    2001-12-01

    The post-irradiation examination of irradiated DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) simulated fuel in HANARO was performed at IMEF (Irradiated Material Examination Facility) in KAERI during 6 months from October 1999 to March 2000. The objectives of this post-irradiation test are i) the integrity of the capsule to be used for DUPIC fuel, ii) ensuring the irradiation requirements of DUPIC fuel at HANARO, iii) performance verification in-core behavior at HANARO of DUPIC simulated fuel, iv) establishing and improvement the data base for DUPIC fuel performance verification codes, and v) establishing the irradiation procedure in HANARO for DUPIC fuel. The post-irradiation examination performed are γ-scanning, profilometry, density, hardness, observation the microstructure and fission product distribution by optical microscope and electron probe microanalyser (EPMA)

  8. Assessment of martensitic steels for advanced fusion reactors

    International Nuclear Information System (INIS)

    Wareing, J.; Tavassoli, A.A.

    1995-01-01

    Martensitic steels are currently considered in Europe to be prime structural candidate materials for the first wall and breeding blanket of the DEMO fusion reactor. In this design, reactor power and wall loading will be significantly higher than those of an experimental reactor. ITER and will give rise to component operating temperatures in the range 250 to 550 0 C with neutron doses higher than 70 dpa. These conditions render austenitic stainless steel, which will be used in ITER, less favourable. Factors contributing to the promotion of martensitic steels are their excellent resistance to irradiation induced swelling, low thermal expansion and high thermal conductivity allied to advanced industrial maturity, compared to other candidate materials vanadium alloys. This paper described the development and optimisation of the steel and weld metal. Using data design rules generated on modified 9 Cr 1 Mo steel during its qualification as a steam generator material for the European Fast Reactor (EFR), interim design guidelines are formulated. Whilst the merits of the steel are validated, it is shown that irradiation embrittlement at low temperature, allied to the need for prolonged post-weld hat treatment and the long term creep response of welds remain areas of some concern. (author). 18 refs., 6 figs., 2 tabs

  9. Remoción de fósforo de diferentes aguas residuales en reactores aeróbios de lecho fluidizado trifásico con circulación interna

    Directory of Open Access Journals (Sweden)

    Gleyce Teixeira Correia

    2013-01-01

    Full Text Available El vertimiento de aguas residuales (AR produce impactos sobre los cuerpos de agua receptores. Nutrientes como P generan implicaciones en los sistemas lénticos pues aceleran los procesos de eutrofización. Se han utilizado diversas tecnologías para la remoción de P de las AR: sistemas de tratamiento físico químico con importantes efectos por adición de productos coagulantes; procesos biológicos basados en alternancia de condiciones anaerobias y aerobias con importantes implicaciones de volumen necesario; sistemas como lagunas de estabilización e irrigación requieren de áreas muy considerables y procesos de postratamiento. Los reactores aerobios de lecho fluidizado con circulación interna (RALFCI son opciones compactas que utilizan gran concentración de biomasa activa que han demostrado su capacidad para remover materia orgánica y N. Para AR domésticas provenientes de la estación de bombeo de Ilha Solteira y para los efluentes de un sistema de recirculación acuícola (SRA de cultivo semi-intensivo de tilapia se evaluó la eficiencia de remoción de P reactivo y P total en tres tipos de RALFCI con diámetro externo de 250 mm y diferentes diámetros de tubo interno (DTI, con dos medios de soporte y diferentes concentraciones en dos de los reactores. Las eficiencias medias de remoción de P reactivo en AR domésticas para un tiempo de retención hidráulica (TRH de 3 horas en el reactor con DTI 125 mm variaron entre 25,6 y 38,4% y en el reactor con DTI 150 mm entre 27,5 y 32,5%; la remoción de P total en el SRA para un TRH de 0,19 h y DTI 100 mm fue de 32,7%.

  10. Some results on development, irradiation and post-irradiation examinations of fuels for fast reactor-actinide burner (MOX and inert matrix fuel)

    International Nuclear Information System (INIS)

    Poplavsky, V.; Zabudko, L.; Moseev, L.; Rogozkin, B.; Kurina, I.

    1996-01-01

    Studies performed have shown principal feasibility of the BN-600 and BN-800 cores to achieve high efficiency of Pu burning when MOX fuel with Pu content up to 45% is used. Valuable experience on irradiation behaviour of oxide fuel with high Pu content (100%) was gained as a result of operation of two BR-10 core loadings where the maximum burnup 14 at.% was reached. Post-irradiation examination (PIE) allowed to reveal some specific features of the fuel with high plutonium content. Principal irradiation and PIE results are presented in the paper. Use of new fuel without U-238 provides the maximum burning capability as in this case the conversion ratio is reduced to zero. Technological investigations of inert matrix fuels have been continued now. Zirconium carbide, zirconium nitride, magnesium oxide and other matrix materials are under consideration. Inert matrices selection criteria are discussed in the paper. Results of technological study, of irradiation in the BOR-60 reactor and PIE results of some inert matrix fuels are summarized in this report. (author). 2 refs, 1 fig., 3 tabs

  11. Generation IV reactors: reactor concepts

    International Nuclear Information System (INIS)

    Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

    2003-01-01

    Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

  12. Calibration of the hydraulic model of a full-scale activated sludge plant; Calibracion hidraulica a escala real de un reactor de lodos activados

    Energy Technology Data Exchange (ETDEWEB)

    Fall, Cheikh [Universidad Autonoma del Estado de Mexico (Mexico); Loaiza-Navia, Jimmy [Servicios de Agua y Drenaje de Monterrey (Mexico)

    2008-04-15

    When planning to simulate a wastewater treatment plant (WWTP) with the activated sludge model number 1 (ASM1), one of the first requirements is to determine the hydraulic model of the reactor. The aim of this study was to evaluate the hydrodynamic regime of the aeration tank of a municipal WWTP by using a rhodamine tracer test and the Aquasim simulation software. A pre-simulation was performed in order to quantify the appropriate colorant mass, set up a sampling plan and evaluate the anticipated visual impact of the tracer test in the river receiving the treated effluents. A tracer test and dynamic flow measurements were carried out, the results of which served to establish and calibrate the hydraulic model. The evaluated tank was physically built as a plug-flow reactor subdivided in 7 compartments, but the study revealed that it is best represented by a model with 5 virtual mixed reactors in series. Through the study, the approach of using a WWTP simulator for hydraulics calibration was shown to be a powerful and flexible tool for designing a tracer test and for identifying adequate tank-in-series models of full-scale activated sludge aeration tanks. [Spanish] Cuando se planea simular una planta de tratamiento con base en el modelo numero 1 de lodos activados (ASM1), uno de los primeros requisitos es determinar el modelo hidraulico del reactor. En este trabajo se estudio el regimen hidrodinamico del tanque de accion de una planta de tratamiento de aguas residuales municipales (PTAR), utilizando una prueba de trazador con rodamina y un programa de simulacion (Aquasim). Se realizo una prueba de trazador con el experimento, lo que permitio determinar la cantidad requerida de trazador, fijar los intervalos de muestreo y limitar el impacto visual anticipado del colorante sobre el rio que recibe el efluente tratado. Se llevaron a cabo la prueba de trazador y la medicion de los perfiles dinamicos de caudales, cuyos resultados sirvieron para establecer y calibrar el

  13. Guidebook on destructive examination of water reactor fuel

    International Nuclear Information System (INIS)

    1997-01-01

    As a result of common efforts of fuel vendors, utilities and research institutes the average burnup pf design batch fuels was increased for both PWRs and BWRs and the fuel failure rate has been reduced. The previously published Guidebook on Non-Destructive Examination of Water Reactor Fuel recommended that more detailed destructive techniques are required for complete understanding of fuel performance. On the basis of contributions of the 14 participants in the ED-WARF-II CRP and proceedings of IAEA Technical Committee on Recent Developments in Post-irradiation Examination Techniques for Water Reactor Fuel this guidebook was compiled. It gives a complete survey of destructive techniques available to date worldwide. The following examination techniques are described in detailed including major principles of equipment design: microstructural studies; elemental analysis; isotopic analysis; measurement of physical properties; measurement of mechanical properties. Besides the examination techniques, methods for refabrication of experimental rods from high burnup power reactor rods as well as methods for verification of non-destructive techniques by using destructive techniques is included

  14. In-pile measurements and PCI fuel modelling of WWER reactors

    International Nuclear Information System (INIS)

    Krett, V.; Novak, J.; Pazdera, F.; Smid, J.

    1984-01-01

    Summary information concerning development of the CEFEUS modular code for the fuel element reliability evaluation is presented in the paper. A concise description of particular modules connected with appropriate experiments is given. The results and aims of irradiation experiments with light water reactor diagnostic assemblies and the post-irradiation examination programme of these assemblies are also briefly discussed. (author)

  15. High temperature fast reactor for hydrogen production in Brazil; Reator nuclear rapido de altissima temperatura para producao de hidrogenio no Brasil

    Energy Technology Data Exchange (ETDEWEB)

    Nascimento, Jamil A. do; Ono, Shizuca; Guimaraes, Lamartine N.F. [Centro Tecnico Aeroespacial (CTA-IEAv), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados]. E-mail: jamil@ieav.cta.br

    2008-07-01

    The main nuclear reactors technology for the Generation IV, on development phase for utilization after 2030, is the fast reactor type with high temperature output to improve the efficiency of the thermo-electric conversion process and to enable applications of the generated heat in industrial process. Currently, water electrolysis and thermo chemical cycles using very high temperature are studied for large scale and long-term hydrogen production, in the future. With the possible oil scarcity and price rise, and the global warming, this application can play an important role in the changes of the world energy matrix. In this context, it is proposed a fast reactor with very high output temperature, {approx} 1000 deg C. This reactor will have a closed fuel cycle; it will be cooled by lead and loaded with nitride fuel. This reactor may be used for hydrogen, heat and electricity production in Brazil. It is discussed a development strategy of the necessary technologies and some important problems are commented. The proposed concept presents characteristics that meet the requirements of the Generation IV reactor class. (author)

  16. Defects investigation in neutron irradiated reactor steels by positron annihilation

    International Nuclear Information System (INIS)

    Slugen, V.

    2003-01-01

    Positron annihilation spectroscopy (PAS) based on positron lifetime measurements using the Pulsed Low Energy Positron System (PLEPS) was applied to the investigation of defects of irradiated and thermally treated reactor pressure vessel (RPV) steels. PLEPS results showed that the changes in microstructure of the RPV-steel properties caused by neutron irradiation and post-irradiation heat treatment can be well detected. From the lifetime measurements in the near-surface region (20-550 nm) the defect density in Russian types of RPV-steels was calculated using the diffusion trapping model. The post-irradiation heat treatment studies performed on non-irradiated specimens are also presented. (author)

  17. Tratamiento de las excretas de cerdo mediante un reactor anaeróbico SCFBR a nivel de banco Treatment of pig excreta using an SCFBR anaerobic reactor

    Directory of Open Access Journals (Sweden)

    Caicedo Luis A.

    1999-06-01

    Full Text Available Un nuevo reactor anaeróbico denominado Sludge Central Fixed Bed Reactor (SCFBR fue construido y evaluado para tratar los residuos líquidos de las granjas porcícolas. El SCFBR está constituido por tres zonas principales. Una zona inferior de lodos, seguida por un módulo empacado ubicado en forma concéntrica y, en la parte superior, una zona de separación sólido, líquido y gas. El reactor de 28,5 1 de volumen de reacción fue evaluado durante 210 días para tres cargas orgánicas de 0,548, 0,421 y 1,239 g DQO/ 1 día. El SCFBR fue alimentado inicialmente en forma discontinua con tiempos de retención hidráulicos (TRH de 10 y 10,7 días. Posteriormente el TRH fue disminuido a 3,87 días con una alimentación en continuo. Para las tres cargas orgánicas de 0,548, 0,421 y 1,239 g DQO/1 día se obtuvieron remociones en la demanda química de oxígeno (DQO de 68%, 81% y 73% y en los sólidos volátiles (SV, de 53,5%, 55,8% y 50,1%, respectivamente. El SCFBR presentó un buen desempeño, re-presentado en las eficiencias de remoción y en la estabilidad observada. Se presenta una microfotografía tomada de una muestra de lodo de la zona inferior del SCFBR, observándose una gran presencia de microorganismos del género Methanosaeta (Methanothrix.

    A new anaerobic reactor called the Sludge Central Fixed Bed Reactor (SCFBR was built and evaluated for the treatment of liquid residue from the pig farms. The SCFBR has three main parts. The lower area is for sludge, the middle part consists of a concentrically

  18. Programme and current status of fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    Suita, T.; Oyama, A.

    1977-01-01

    In 1967 the Japan Atomic Energy Commission revised her long term programme after a two year study for giving principles to her nuclear energy development programme, which indicated the dominant role of nuclear energy mid 1980's in the electric power generation and stressed the necessity of developing fast breeder reactors. It also recommended to organize a nucleus to undertake this nation-wide project, bringing together the total capability available throughout the country. Accordingly, the Power Reactor and Nuclear Fuel Development Corporation (PNC) was established in 1967 to develop two sodium-cooled fast reactors, an experimental fast reactor of about 100 MW thermal and a prototype fast breeder reactor of about 300 MW electrical, both using mixed oxide fuels. Construction of the experimental fast reactor started in 1970 and was essentially completed at the end of in 1974. The precommissioning test was followed in parallel with re-evaluating quality assurance of all systems. Physics test will be initiated around the end of 1976. The conceptual design of the prototype fast breeder reactor is now toward its final stage. Surveys on its proposed site have just started. Construction will start in 1978. Beside R and D works conducted by many organizations in Japan as well as under the international cooperation, several key test facilities were installed by PNC itself to conduct in-sodium test of full-size prototype components including 50 MW steam generators and post-irradiation-examination of fuels and materials. Recently an interim report was issued to an ad-hoc committee organized by JAEC to evaluate future prospect of the fuel cycle and power reactors. This recommended start of construction of the prototype reactor as scheduled and the large demonstration reactor to be followed to the prototype. Thus the fast breeder reactor is indicated as the most indispensable in 1990's

  19. Research reactor fuel development at AECL

    International Nuclear Information System (INIS)

    Sears, D.F.; Wang, N.

    2000-09-01

    This paper reviews recent U 3 Si 2 and U-Mo dispersion fuel development activities at AECL. The scope of work includes fabrication development, irradiation testing, post-irradiation examination and performance qualification. U-Mo alloys with a variety of compositions, ranging from 6 to 10 wt % Mo, have been fabricated with high purity and homogeneity in the product. The alloys and powders were characterized using optical and scanning electron microscopy, chemical analysis, and X-ray diffraction and neutron diffraction analysis. U-Mo powder samples have been supplied to the Argonne National Laboratory for irradiation testing in the ATR reactor. Low-enriched uranium fuel elements containing U-7 wt % Mo and U-10 wt % Mo with loadings up to 4.5 gU/cm 3 have been fabricated at CRL for irradiation testing in the NRU reactor. The U-Mo fuel elements will be tested in NRU at linear powers up to 145 kW/m, and to 85 atom % 235 U burnup. (author)

  20. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

    Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

  1. The CAREM reactor and present currents in reactor design

    International Nuclear Information System (INIS)

    Ordonez, J.P.

    1990-01-01

    INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author) [es

  2. The advanced test reactor national scientific user facility advancing nuclear technology

    International Nuclear Information System (INIS)

    Allen, T.R.; Thelen, M.C.; Meyer, M.K.; Marshall, F.M.; Foster, J.; Benson, J.B.

    2009-01-01

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  3. Development of a 3-D flow analysis computer program for integral reactor

    International Nuclear Information System (INIS)

    Youn, H. Y.; Lee, K. H.; Kim, H. K.; Whang, Y. D.; Kim, H. C.

    2003-01-01

    A 3-D computational fluid dynamics program TASS-3D is being developed for the flow analysis of primary coolant system consists of complex geometries such as SMART. A pre/post processor also is being developed to reduce the pre/post processing works such as a computational grid generation, set-up the analysis conditions and analysis of the calculated results. TASS-3D solver employs a non-orthogonal coordinate system and FVM based on the non-staggered grid system. The program includes the various models to simulate the physical phenomena expected to be occurred in the integral reactor and will be coupled with core dynamics code, core T/H code and the secondary system code modules. Currently, the application of TASS-3D is limited to the single phase of liquid, but the code will be further developed including 2-phase phenomena expected for the normal operation and the various transients of the integrator reactor in the next stage

  4. Neutron behavior, reactor control, and reactor heat transfer. Volume four

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

  5. Validación de la limpieza del reactor empleado en la preparación de medicamentos

    Directory of Open Access Journals (Sweden)

    Lisseux Castilla Valentín

    2001-04-01

    Full Text Available En la actualidad, la validación de los procesos se ha convertido en una imperiosa necesidad de la Industria Médico-Farmacéutica, para garantizar la calidad de sus productos y lograr la comercialización de estos. En esta dirección se comenzaron los trabajos de validación de la limpieza de los equipos y se realizó la validación de la limpieza del reactor SEN, que se utiliza para la preparación de los inyectables. Para lograr este objetivo, se elaboró una metodología de trabajo, se seleccionaron los métodos analíticos más apropiados, se establecieron los criterios de aceptación y se escribió un protocolo de validación, que constituyó la herramienta fundamental de trabajo. Posteriormente se realizó la validación de la limpieza del reactor, y se concluyó que el procedimiento de limpieza, aunque no garantiza total eliminación de los residuos del producto, sí cumple con los criterios de aceptación establecidos.At present, the validation of the processes is an imperative necessity of the Medicopharmaceutical Industry to guarantee the quality of its products and to commercialize them. To this end, the cleaning of the equipment began to be validated and the validation of the cleaning of the SEN reactor that is used for the preparation of injections was carried out. To attain this goal, a working methodology was designed, the most appropiate analytical methods were selected, the acceptance criteria were established and a validation protocol was written that was the fundamental working tool. Later on, the validation of the cleaning of the reactor was made and it was concluded that although the cleaning procedure does not guarantee the total elimination of the residuals of the product, it fulfills the established acceptance criteria.

  6. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  7. Reactor science and technology: operation and control of reactors

    International Nuclear Information System (INIS)

    Qiu Junlong

    1994-01-01

    This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

  8. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  9. Effects of mecchanical loads due to power excursions on the reactor tank

    International Nuclear Information System (INIS)

    Meier, S.

    1982-06-01

    Coupled fluid dynamics/structural mechanics codes are developed since 10 years to solve problems in the field of reactor safety. Experimental programmes devised to validate these codes should include scaled models that closely resemble real reactor geometries. During tests with these models, fluid movements as well as the structural strains should be comparable to those arising in the reactor tank under accident conditions. The second shot in the 1/6 scaled model of the SNR-300 fits these conditions. The SEURBNUK post shot calculation demonstrates the capability of the code with adequate results for all salient physical values. But the experiment and consequently the calculation is for validation purposes only suited in a limited way because of the uncertainty of the charge behaviour during the shot. (orig.) [de

  10. Design and construction of the SIPPING for fuels of the TRIGA Mark III reactor; Diseno y construccion del SIPPING para combustibles del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Castaneda J, G.; Delfin L, A.; Alvarado P, R.; Mazon R, R.; Ortega V, B. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: adl@nuclear.inin.mx

    2003-07-01

    The sipping technique, it has been used by several possessors of nuclear research reactors in its irradiated nuclear fuels, likewise in some fuel storage sites, with the objective of to determine the quantity of radioactivity that the fuel liberates in the means in that it is. The irradiated fuel in storage of some nuclear research reactors, its can have cracks that cross the cladding of the same one, generating the liberation of fission products that its need to determine to maintain safety measures appropriate as much as the fuel as of the facilities where they are. It doesn't exist until now, some method published for the non destructive sipping test technique. Based on that described, the Reactor Department of the National Institute of Nuclear Research, it has designed and built an inspection system of irradiated fuel that it will allow the detection of gassy fission products in site, and solids by means of the measurement of the activity of the Cs-137 contained in water samples. (Author)

  11. Nuclear data needs for the analysis of generation and burn-up of actinide isotopes in nuclear reactors

    International Nuclear Information System (INIS)

    Kuesters, H.

    1980-04-01

    A reliable prediction of the in-pile and out-of-pile physics characteristics of nuclear fuel is one of the objectives of present-day reactor physics. The paper describes the main production paths of important actinides for light water and fast breeder reactors. The accuracy of recent nuclear data is examined by comparisons of theoretical predictions with the results from post-irradiation analysis of nuclear fuel from power reactors, and partly with results obtained in zero-power facilities. A world-wide comparison of nuclear data to be used in large fast power reactor burn-up and long term considerations is presented. The needs for further improvement of nuclear data are discussed. (orig.) [de

  12. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

    This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

  13. Micropollutant removal from black water and grey water sludge in a UASB-GAC reactor.

    Science.gov (United States)

    Butkovskyi, A; Sevenou, L; Meulepas, R J W; Hernandez Leal, L; Zeeman, G; Rijnaarts, H H M

    2018-02-01

    The effect of granular activated carbon (GAC) addition on the removal of diclofenac, ibuprofen, metoprolol, galaxolide and triclosan in a up-flow anaerobic sludge blanket (UASB) reactor was studied. Prior to the reactor studies, batch experiments indicated that addition of activated carbon to UASB sludge can decrease micropollutant concentrations in both liquid phase and sludge. In continuous experiments, two UASB reactors were operated for 260 days at an HRT of 20 days, using a mixture of source separated black water and sludge from aerobic grey water treatment as influent. GAC (5.7 g per liter of reactor volume) was added to one of the reactors on day 138. No significant difference in COD removal and biogas production between reactors with and without GAC addition was observed. In the presence of GAC, fewer micropollutants were washed out with the effluent and a lower accumulation of micropollutants in sludge and particulate organic matter occurred, which is an advantage in micropollutant emission reduction from wastewater. However, the removal of micropollutants by adding GAC to a UASB reactor would require more activated carbon compared to effluent post-treatment. Additional research is needed to estimate the effect of bioregeneration on the lifetime of activated carbon in a UASB-GAC reactor.

  14. Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

    International Nuclear Information System (INIS)

    Hirayama, Shoichi; Kawabe, Takaya

    1995-01-01

    The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

  15. MID-VASTUS VS MEDIAL PARA-PATELLAR APPROACH IN TOTAL KNEE REPLACEMENT—TIME TO DISCHARGE

    Science.gov (United States)

    Mukherjee, P.; Press, J.; Hockings, M.

    2009-01-01

    Background It has been shown before that when compared with the medial para-patellar approach, the mid-vastus approach for TKR results in less post-operative pain for patients and more rapid recovery of straight leg raise. As far as we are aware the post-operative length of stay of the two groups of patients has not been compared. We postulated that the reduced pain and more rapid recovery of straight leg raise would translate into an earlier, safe, discharge home for the mid-vastus patients compared with those who underwent a traditional medial para-patellar approach. Methods Twenty patients operated on by each of five established knee arthroplasty surgeons were evaluated prospectively with regard to their pre and post-operative range of movement, time to achieve straight leg raise post-operatively and length of post-operative hospital stay. Only one of the surgeons performed the mid-vastus approach, and the measurements were recorded by physiotherapists who were blinded as to the approach used on each patient. Results The results were analysed using a standard statistical software package, and although the mean length of stay was lower for the mid-vastus patients, the difference did not reach a level of significance (p = 0.13). The time taken to achieve straight leg raise post-operatively was significantly less in the mid-vastus group (p<0.001). Conclusion Although this study confirms previous findings that the mid-vastus approach reduces the time taken for patients to achieve straight leg raise, when compared with the medial para-patellar approach, on its own it does not translate into a significantly shorter length of hospital stay. In order to reduce the length of post-operative hospital stay with an accelerated rehabilitation program for TKR, a multi-disciplinary approach is required. Patient expectations, GP support, physiotherapists and nursing staff all have a role to play and the mid-vastus approach, in permitting earlier straight leg raising

  16. Fuel supply demand balances for future FBR commercialization: impacts on plutonium pricing and reactor design

    International Nuclear Information System (INIS)

    Braun, C.; Zebroski, E.L.

    1985-01-01

    Plutonium supply and demand balances for fast breeder reactor (FBR) commercialization post-2000 were computed to determine: (a) the maximum supportable number of FBRs that could be installed based on plutonium availability considerations and (b) the feasibility of a reasonable FBR capacity growth case assuming slow introduction post-2010 and rapid capacity growth post-2035. The purpose of the analysis was to determine the outer limitation on the maximum future FBR introduction, or the bounds of a possible plutonium-limited introduction rate, and to estimate the reasonableness of a more limited capacity growth case

  17. Reactor physics aspects of CANDU reactors

    International Nuclear Information System (INIS)

    Critoph, E.

    1980-01-01

    These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

  18. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  19. Comparison between TRU burning reactors and commercial fast reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

    2001-03-01

    Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

  20. Análisis de la dispersión axial de masa y calor en reactores de lecho fijo

    Directory of Open Access Journals (Sweden)

    Rangel Jara Hermes Augusto

    1997-01-01

    Full Text Available Dentro del espíritu investigativo a nivel teórico del estudio de los reactores químicos, el presente trabajo desarrolla e implementa un análisis conceptual y numérico de los fenómenos de dispersión axial de calor y masa en reactores de lecho fijo. Se pretende disponer de una alternativa numérica que permita en una forma rápida y precisa la solución de las ecuaciones diferenciales junto con las respectivas condiciones de frontera del modelo matemático. Para la simulación del reactor de lecho fijo se empleó un modelo unidimensional pseudohomogeneo con parámetros aglomerados.

  1. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

  2. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    Kouts, H.

    1986-01-01

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

  3. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  4. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  5. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

  6. Potencial de las imágenes UAV como datos de verdad terreno para la clasificación de la severidad de quema de imágenes Landsat: aproximaciones a un producto útil para la gestión post incendio

    Directory of Open Access Journals (Sweden)

    M. Pla

    2017-12-01

    Full Text Available La cuantificación de la severidad de los incendios forestales es determinante para conocer la evolución del paisaje después de un incendio forestal y provee información de gran utilidad frente a la toma de decisiones en la gestión post incendio. La cartografía cuantitativa de severidad de incendios a partir de cambios relativos del índice Normalized Burn Ratio (RdNBR no está siendo realmente incorporada en los procesos de toma de decisiones, siendo más utilizada la categorización en niveles de severidad (alta, mediana y baja. Sin embargo, las clasificaciones de severidad más comunes, basadas en la definición de umbrales de corte de RdNBR a partir de información de campo, no son siempre posibles por falta de datos de campo o bien porque los umbrales publicados resultan poco satisfactorios en localizaciones distintas a las de su calibración. El auge del uso de UAVs (Unmaned Aerial Vehicle ha planteado estas plataformas como posible herramienta para la validación de información de satélite. En el presente trabajo se presenta la potencialidad de los UAV como información de verdad terreno en incendios forestales. A partir de la fotointerpretación de imágenes RGB de alta resolución se ha creado el índice ASPI (Aerial Severity Proportion Index, el cual, a partir de modelos de regresión no lineales con el índice RdNBR, permite delimitar umbrales para la clasificación de las imágenes Landsat y obtener un mapa cualitativo de severidad. La validación de los modelos de regresión entre RdNDR y ASPI a partir de puntos al azar muestra un índice kappa de 0,5 con un acierto relativo del 70,8%. Por lo tanto, las imágenes UAV son una herramienta muy útil para la clasificación de la severidad de incendios forestales y para rellenar la brecha existente entre la información proveniente de imágenes de satélite y las costosas campañas de campo.

  7. Post-Construction Testing of the Elk River, Hallam and Piqua Power Reactor Plants; Essais apres construction des centrales nucleaires d'Elk River, de Hallam et de Piqua; Predehkspluatatsionnoe ispytanie Ehlk-riverskoj, Khehlpemskoj i Pikuaskoj ehnergeticheskikh reaktornykh ustanovok; Ensayos posteriores a la construccion de las centrales nucleoelectricas de Elk River, Hallam y Piqua

    Energy Technology Data Exchange (ETDEWEB)

    Pursel, C. A. [United States Atomic Energy Commission, Argonne, IL (United States)

    1963-10-15

    Actual experience gained during the post-construction testing of three nuclear power plants, under the USAEC Power Reactor Demonstration Program, may permit some generalizations concerning this phase of plant construction and operation. The three plants, Elk River Reactor (ERR), Hallam Nuclear Power Facility (HNPF), and the Piqua Nuclear Power Facility (PNPF), represent three different reactor concepts: natural-circulation boiling water, sodiumgraphite, and organic cooled and moderated, respectively. The post-construction testing period included the time between the end of construction (erection of structures and installation of equipment) and the beginning of power operation (generation of significant net electrical power). The tests were intended to: (a) verify the performance characteristics of the as-installed equipment; (b) obtain initial criticality and reactivity coefficient measurements; and (c) determine reactor physics and plant performance characteristics at a sequence of increasing power levels. .The experience gained can be reported in six separate but interrelated categories: (1) schedule; (2) costs; (3) staffing requirements; (4) procedures; (5) equipment performance (including malfunctions); and (6) actual, as compared to predicted, system performance characteristics. The average project staffing, including craftsmen, operators, supervisors, technical support and trainees, was approximately 50 for ERR, 115 for HNPF, and 60 for PNPF. Detailed written Pre-operational Test Procedures were prepared for each major component and system. To the maximum possible extent, all tests were performed before fuel loading and operation of the integrated plant. Authorization procedures (duplicates of the licensing procedures for non-USAEC-owned plants) were in progress during almost all of the post-construction testing periods. The time required for post-construction testing of each of these plants significantly exceeded the original estimates. The tests disclosed

  8. The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

    International Nuclear Information System (INIS)

    Allen, T.R.; Benson, J.B.; Foster, J.A.; Marshall, F.M.; Meyer, M.K.; Thelen, M.C.

    2009-01-01

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  9. Modelamiento matemático para la pirolisis del cuesco de palma aceitera- Mathematical Modelingfor the Pyrolysis of the Oil Palm Kernel Shell

    Directory of Open Access Journals (Sweden)

    Manuel Alejandro Mayorga Betancourt

    2017-06-01

    Full Text Available En Colombia hay diferentes procesos en los cuales se desperdician los residuos, uno de ellos es la obtención de aceite para producir biodiesel a partir de palma aceitera, proceso en el cual se generan importantes cantidades de cuesco de palma, siendo los procesos termoquímicos una de las formas de aprovechamiento energético. El resultado del presente trabajo, es proponer un modelo matemático para el comportamiento del cuesco de palma africana en el proceso de pirolisis y en un estudio posterior para la gasificación. Este trabajo fue desarrollado haciendo una descripción del proceso en un reactor de lecho fijo tubular, el cual se utiliza para ambos procesos, tanto pirolisis como gasificación, con calentamiento directo, siendo muy exotérmico el proceso, identificando la fenomenología en la cual se aplican los conceptos de transferencia de energía, masa. Para el proceso de pirolisis se plantearon los balances de transferencia de energía y masa, despreciando el balance de momento debido a que los gases de síntesis se retiran para que no se generen reacciones heterogéneas, eliminando la fase gaseosa para no tener caídas de presión, por lo cual solo se tuvo en cuenta la fase sólida, lo que permitió que se tratara como una reacción homogénea. Como resultados se generaron dos modelos que describen el comportamiento del reactor en el proceso de pirolisis como un paso inicial para contribuir a la estandarización del proceso a nivel industrial.

  10. Comparación de la técnica de Dennis con los hallazgos hepáticos post - mortem para el diagnóstico de la fasciolosis bovina

    Directory of Open Access Journals (Sweden)

    Alejandra Alvarez

    2009-10-01

    Full Text Available El presente estudio hace una comparación entre los resultados de la técnica de Dennis para diagnóstico de Fasciolosis bovina, frente a hallazgos post mortem, en hígados de bovinos faenados en la empresa Matadero de Tunja. Es un estudio experimental descriptivo, por cuanto expone los hallazgos en matadero frente a los arrojados por el método de Dennis. La población total de bovinos adultos fue de 2800, de los cuales se tomó una muestra de 139 animales. El muestreo se llevó a cabo durante el mes de febrero del año 2009. Los resultados indican que la técnica de Dennis no es lo suficientemente sensible para el diagnóstico de Fasciola hepática, por lo que se debe evaluar su uso rutinario.La técnica coprológica se fundamenta en el principio de sedimentación delos huevos, por lo tanto depende la salida de éstos en la materia fecal, lo que la hace poco efectiva para la detección del parásito. La razón de este estudiose sustenta en que la enfermedad se encuentra ampliamente distribuida en la región y en el país. Su diagnóstico se realiza rutinariamente mediante técnicas coprológicas aplicadas a los animales en los que se sospecha la enfermedad.

  11. De redes sociales recíprocas a grupos de acción para el intercambio de mercado: la “privatización espontánea” en la Hungría post-comunista

    Directory of Open Access Journals (Sweden)

    Larissa Lomnitz

    2011-12-01

    Full Text Available Siguiendo el trabajo previo sobre la importancia que han tenido las redes sociales para la supervivencia económica y social del funcionariado de clase media latinoamericano y soviético, este artículo explora el papel de las redes sociales (las conexiones en el proceso de privatización y liberalización del mercado en la Hungría post-comunista. Nos basamos en estudios académicos precedentes y en trabajo de campo desarrollado durante varios meses en Budapest para mostrar que las redes sociales son estructuras intermediarias centrales en las que los individuos y los grupos construyen soluciones que les permiten sobrellevar las deficiencias del sistema formal. Desde esta perspectiva, exploraremos la importancia de las conexiones entre gerentes durante el primer periodo de privatizaciones en Hungría, conocido como “privatización espontanea”.

  12. Available post-irradiation examination techniques at Romanian institute for nuclear research

    International Nuclear Information System (INIS)

    Parvan, Marcel; Sorescu, Antonius; Mincu, Marin; Uta, Octavian; Dobrin, Relu

    2005-01-01

    The Romanian Institute for Nuclear Research (INR) has a set of nuclear facilities consisting of TRIGA 14 MW(th) materials testing reactor and LEPI (Romanian acronym for post-irradiation examination laboratory) which enable to investigate the behaviour of the nuclear fuel and materials under various irradiation conditions. The available techniques of post-irradiation examination (PIE) and purposes of PIE for CANDU reactor fuel are as follows. 1) Visual inspection and photography by periscope: To examine the surface condition such as deposits, corrosion etc. 2) Eddy current testing: To verify the cladding integrity. 3) Profilometry and length measurement performed both before and after irradiation: To measure the parameters which highlight the dimensional changes i.e. diameter, length, diametral and axial sheath deformation, circumferential sheath ridging height, bow and ovality. 4) Gamma scanning and Tomography: To determine the burnup, axial and radial fission products activity distribution and to check for flux peaking and loading homogeneity. 5) Puncture test: To measure the pressure, volume and composition of fission gas and the inner free volume. 6) Optical microscopy: To highlight the structural changes and hydriding, to examine the condition of the fuel-sheath interface and to measure the oxide thickness and Vickers microhardness. 7) Mass spectrometry: To measure the burnup. 8) Tensile testing: To check the mechanical properties. So far, non-destructive and destructive post-irradiation examinations have been performed on a significant number of CANDU fuel rods (about 100) manufactured by INR and irradiated to different power histories in the INR 14 MW(th) TRIGA reactor. These examinations have been performed as part of the Romanian research programme for the manufacturing, development and safety of the CANDU fuel. The paper describes the PIE techniques and some results. (Author)

  13. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  14. Final project report: TA-35 Los Alamos Power Reactor Experiment No. II (LAPRE II) decommissioning project

    International Nuclear Information System (INIS)

    Montoya, G.M.

    1993-02-01

    This final report addresses the decommissioning of the LAPRE II Reactor, safety enclosure, fuel reservoir tanks, emergency fuel recovery system, primary pump pit, secondary loop, associated piping, and the post-remediation activities. Post-remedial action measurements are also included. The cost of the project including, Phase I assessment and Phase II remediation was approximately $496K. The decommissioning operation produced 533 M 3 of mixed waste

  15. Behavior of actinides in the Integral Fast Reactor fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Lineberry, M.J. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

    1994-06-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors` confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

  16. Actinide behavior in the Integral Fast Reactor. Final project report

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C.

    1994-11-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction rates and neutron spectra. These experimental data increase the authors confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

  17. Actinide behavior in the Integral Fast Reactor. Final project report

    International Nuclear Information System (INIS)

    Courtney, J.C.

    1994-11-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ( 237 Np, 240 Pu, 241 Am, and 243 Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction rates and neutron spectra. These experimental data increase the authors confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs

  18. Behavior of actinides in the Integral Fast Reactor fuel cycle

    International Nuclear Information System (INIS)

    Courtney, J.C.; Lineberry, M.J.

    1994-01-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ( 237 Np, 240 Pu, 241 Am, and 243 Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors' confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs

  19. In-pile critical heat flux and post-dryout heat transfer measurements – A historical perspective

    Energy Technology Data Exchange (ETDEWEB)

    Groeneveld, D.C., E-mail: degroeneveld@gmail.com

    2017-06-15

    In the 1960s’ and 1970s’ Canada was a world leader in performing in-reactor heat transfer experiments on fuel bundles instrumented with miniature sheath thermocouples. Several Critical Heat Flux (CHF) and Post-CHF experiments were performed in Chalk River’s NRU and NRX reactors on water-cooled 3-, 18-, 19-, 21-, and 36-element fuel bundles. Most experiments were obtained at steady-state conditions, where the power was raised gradually from single-phase conditions up to the CHF and beyond. Occasionally, post-dryout temperatures up to 600 °C were maintained for several hours. In some tests, the fuel behaviour during loss-of-flow and blowdown transients was investigated – during these transients sheath temperatures could exceed 2000 °C. Because of the increasingly more stringent licensing requirements for in-pile heat transfer tests on instrumented fuel bundles, no in-pile CHF and post-dryout tests on fuel bundles have been performed anywhere in the world for the past 40 years. This paper provides details of these unique in-pile experiments and describes some of their heat transfer results.

  20. Survey of post-irradiation examinations made of mixed carbide fuels

    International Nuclear Information System (INIS)

    Coquerelle, M.

    1997-01-01

    Post-irradiation examinations on mixed carbide, nitride and carbonitride fuels irradiated in fast flux reactors Rapsodie and DFR were carried out during the seventies and early eighties. In this report, emphasis was put on the fission gas release, cladding carburization and head-end gaseous oxidation process of these fuels, in particular, of mixed carbides. (author). 8 refs, 16 figs, 3 tabs

  1. Study of a fuel assembly for the nuclear reactor of IV generation cooled with supercritical water; Estudio de un ensamble de combustible para el reactor nuclear de generacion IV enfriado con agua supercritica

    Energy Technology Data Exchange (ETDEWEB)

    Barragan M, A.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Espinosa P, G., E-mail: albrm29@yahoo.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (MX)

    2011-11-15

    In this work a neutron study is presented about a square assembly design of double line of fuel rods, with moderator box to the center of the arrangement, for a nuclear reactor cooled with supercritical water (SCWR). The SCWR reactor was chosen by the characteristics of its design, mainly because is based in light water reactors (PWR and BWR), and the operational experience that has of them allow to use models and similar programs to simulate the fuel and the nucleus of this type of reactors. To develop the necessary models and to carry out the design and analysis of the SCWR reactor, the neutron codes MCNPX and Helios were used. The reason of using both codes, is because the code MCNPX used thoroughly in the neutron simulation of these reactors, it has been our reference code to analyze the results obtained with the Helios code which results are more efficient because its calculation times are minors. In the nucleus design the same parameters for both codes were considered. The results show that the design with Helios is a viable option to simulate these reactors since their values of the neutrons multiplication factor are very similar to those obtained with MCNPX. On the other hand, it could be corroborated that the CASMO-4 code is inadequate to simulate the fuel to the temperature conditions and water pressure in the SCWR. (Author)

  2. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  3. Advanced In-Core Fuel Cycles for the Gas Turbine-Modular Helium Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto

    2006-04-15

    Amid generation IV of nuclear power plants, the Gas Turbine - Modular Helium Reactor, designed by General Atomics, is the only core with an energy conversion efficiency of 50%; the safety aspects, coupled to construction and operation costs lower than ordinary Light Water Reactors, renders the Gas Turbine - Modular Helium reactor rather unequaled. In the present studies we investigated the possibility to operate the GT-MHR with two types of fuels: LWRs waste and thorium; since thorium is made of only fertile {sup 232}Th, we tried to mix it with pure {sup 233}U, {sup 235}U or {sup 239}Pu; ex post facto, only uranium isotopes allow the reactor operation, that induced us to examine the possibility to use a mixture of uranium, enriched 20% in {sup 235}U, and thorium. We performed all calculations by the MCNP and MCB codes, which allowed to model the reactor in a very detailed three-dimensional geometry and to describe the nuclides transmutation in a continuous energy approach; finally, we completed our studies by verifying the influence of the major nuclear data libraries, JEFF, JENDL and ENDF/B, on the obtained results.

  4. The irradiation test program for transmutation in the French Phenix fast reactor

    International Nuclear Information System (INIS)

    Guidez, J.; Chaucheprat, P.; Fontaine, B.; Brunon, E.

    2004-01-01

    Put on commercial operation in July 1974, the French fast reactor Phenix reached a 100 000 hours operation time in september 2003. When the French law relative to long lived radioactive waste management was promulgated on December 1991, priority was given to Phenix to be run as a research reactor and to carry on a wide irradiation program dedicated to study transmutation of minor actinides and long-lived fission products. After a major renovation program required to extend the reactor lifetime, Phenix power buildup took place in 2003. Experimental irradiations have been loaded in the core, involving components for heterogeneous and homogeneous transmutation modes, americium targets, technetium 99 metal pins and isolated isotopes for integral cross-sections measurements. Associated post- irradiated examination programs are already underway or planned. With new experiments to be loaded in the core in 2006 the Phenix reactor remains to be a powerful tool providing an important experimental data on fast reactors and on transmutation of minor actinides and long-lived fission products, as well as it will contribute to gain further experience in the framework of the GENERATION IV International Forum. (authors)

  5. Research reactor standards and their impact on the TRIGA reactor community

    International Nuclear Information System (INIS)

    Richards, W.J.

    1980-01-01

    The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

  6. Long-term recovery of pressurized water reactors following a large break loss-of-coolant accident

    International Nuclear Information System (INIS)

    Fletcher, C.D.; Callow, R.A.

    1989-01-01

    The USNRC recently identified a possible safety concern for PWR's. Following the reflood phase of a large break loss-of-coolant accident, long-term cooling of the reactor core may not be ensured. Specifically, the concern is that, for a pump discharge cold leg break, the loop seals in the reactor coolant pump suction piping will refill with liquid and the post-reflood steam production may depress the liquid levels in the downflow sides of the loop seals. A loop seal depression would cause a corresponding depression of the core liquid levels and possibly a fuel rod heatup in the upper core region. This paper is intended as an introduction of the safety issue that: 1) describes the important aspects of the problem, 2) provides an initial analysis of the consequences, and 3) discusses ongoing work in this area. Because the elevation of the loop seals is near the mid-core elevation in plants of WE design, the concern is greatest for those plants. There is less concern for most plants of CE design, and likely no concern for plants of BW design. This issue was addressed by employing both steady-state and transient systems analysis approaches. Two approaches were used because of uncertainties regarding actual reactor coolant system behavior during the post-reflood period. The steady-state approach involved the development and application of a simple computer program to investigate reactor coolant system behavior assuming quiescent post-reflood conditions. The transient systems approach involved investigating this behavior using the RELAP5/MOD2 computer code and a comprehensive RELAP5 model of a WE PWR. The steady-state analysis indicated only a moderate fuel rod heatup is possible. The transient systems analysis indicated boiling and condensation-induced flow oscillations are sufficient to prevent fuel rod heatup. Analysis uncertainties are discussed. (orig./HP)

  7. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  8. Improved nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

  9. Reactor as furnace and reactor as lamp

    International Nuclear Information System (INIS)

    Goldanskii, V.I.

    1992-01-01

    There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

  10. Nuclear reactor plant with a small gas-cooled HT reactor accommodated in a steel pressure vessel

    International Nuclear Information System (INIS)

    Schoening, J.; Elter, C.

    1986-01-01

    The plant has a small HT reactor and an He/He heat exchanger situated above this, with preferably two parallel circulating blowers connected after it. It also has at least one post-shutdown heat removal system, which is situated after the He/He heat exchanger in the direction of flow and which always has the total quantity of primary helium flowing through it. In one version of the design, the heat exchanger consists of two concentric bundles of helices connected after one another, which have primary helium flowing in one direction and secondary helium in the opposite direction. (orig./HP) [de

  11. Development of Reactor Console Simulator for PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Mohd Idris Taib; Izhar Abu Hussin; Mohd Khairulezwan Abdul Manan; Nufarhana Ayuni Joha; Mohd Sabri Minhat

    2012-01-01

    The Reactor Console Simulator will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behaviour and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of man-machine interface is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate the estimated reactor console parameters. (author)

  12. Studies on the assessment and validation of reactor dynamics models used in Finland

    International Nuclear Information System (INIS)

    Vanttola, T.

    1993-10-01

    Two reactor dynamics related computer codes of the calculation system at the Technical Research Centre of Finland have been assessed. The codes TRAB and SMATRA, have been examined from two points of view. First, models of some critical phenomena determining the worst fuel rod conditions during reactor transients have been evaluated on the basis of experimental information. Second, the the overall behaviour of the codes describing the dynamics of the reactor core and its cooling system has been studied on the basis of simulation of real transients and of performed safety analyses of selected accidents. The emphasis is on the VVER-440 reactors, but the generality of the methods has been demonstrated by showing that the key phenomena of the Chernobyl accident can be reproduced and analysed using the same calculation system. In the study the separate phenomena examined are single- and two-phase friction, post DNB heat transfer and critical heat flux in the VVER rod bundle. (60 refs., 11 figs., 4 tabs.)

  13. The accident at Chernobyl and its implications for the safety of CANDU reactors

    International Nuclear Information System (INIS)

    1987-05-01

    In August 1986, a delegation of Canadians, including two members of the staff of the AECB (Atomic Energy Control Board), attended a post-accident review meeting in Vienna, at which Soviet representatives described the accident and its causes and consequences. On the basis of the information presented at that meeting, AECB staff conducted a study of the accident to ascertain its implications for the safety of CANDU nuclear reactors and for the regulatory process in Canada. The conclusion of this review is that the accident at Chernobyl has not revealed any important new information which would have an effect on the safety requirements for CANDU reactors as presently applied by the AECB. All important aspects of the accident and its causes have been considered by the AECB in the licensing process for currently licensed reactors. However a number of recommendations are made with respect to aspects of reactor safety which should be re-examined in order to reinforce this conclusion

  14. European development of ferritic-martensitic steels for fast reactor wrapper applications

    International Nuclear Information System (INIS)

    Bagley, K.; Little, E.A.; Levy, V.; Alamo, A.

    1987-01-01

    9-12%Cr ferritic-martensitic stainless steels are under development in Europe for fast reactor sub-assembly wrapper applications. Within this class of alloys, attention is focussed on three key specifications, viz. FV448 and DIN 1.4914 (both 10-12%CrMoVNb steels) and EM10 (an 8-10%Cr-0.15%C steel), which can be optimized to give acceptably low ductile-brittle transition characteristics. The results of studies on these steels, and earlier choices, covering heat treatment and compositional optimization, evolution of wrapper fabrication routes, pre and post-irradiation mechanical property and fracture toughness behaviour, microstructural stability, void swelling and in-reactor creep characteristics are reviewed. The retention of high void swelling to displacement doses in excess of 100 dpa in reactor irradiations reaffirms the selection of 9-12%Cr steels for on-going wrapper development. Moreover, irradiation-induced changes in mechanical properties (e.g. in-reactor creep and impact behaviour), measured to intermediate doses, do not give cause for concern; however, additional data to higher doses and at the lower irradiation temperatures of 370 0 -400 0 C are needed in order to fully endorse these alloys for high burnup applications in advanced reactor systems

  15. Nuclear power plant with improved arrangements for the removal of post fission and emergency heating

    International Nuclear Information System (INIS)

    Buescher, E.; Vinzens, K.

    1977-01-01

    This is concerned with additional equipment for emergency heat removal in a sodium cooled reactor, which operates on failure of the post fission heat removal system. The space for pressure relieving spaces and concrete masses as heat sinks within the reactor cell is no longer required. In this nuclear power plant, a heat exchanger chain transmits heat and power: There is a first sodium circuit between pressure vessel and the first heat exchanger, a second one between the first and second heat excahngers, and a third (Steam) circuit with turbine, condenser and return pump. A fourth circuit connects the secondary side of the condenser with a cooling tower. There is a threee component heat excahgner in the primary circuit after the first heat exchanger, which is built around the first heat exchanger, and is sealed into an unloading space. This space is situated next to the reactor cell and is above the operating level of the sodium in the pressure vessel. It is connected to the cell by an upper duct, normally closed by a bursting disc, and by a lower duct. In the three comopnent heat exchanger, a liquid lead-bismuth eutectic mixture transmits the heat from sodium pipes to water pipes. In normal operation it is used for steam superheating or feedwater preheating. The three component heat exchanger bridges the first and second heat exchangers as an emergency heat exchanger. If in such a case the post fission heat removal has failed, the sodium evaporating in the pressure vessel flows into the unloading space and condenses on the ribs of the emergency heat exchanger. The post fission heat is fed by the water secondary medium directly into the tertiary circuit. The sodium condensate flows back from the unloading space via the lower duct into the reactor cell and maintains the emergency level there. (RW) 891 RW [de

  16. Mechanism design for the control rods conduction of TRIGA Mark III reactor in the NINR; Diseno del mecanismo para la conduccion de las barras de control del reactor Triga Mark III del ININ.

    Energy Technology Data Exchange (ETDEWEB)

    Franco C, A

    1997-12-01

    This work presents in the first chapter a general studio about the reactor and the importance of control rods in the reactor , the mechaniucal design attending to requisitions that are imposed for conditions of operation of the reactor are present in the second chapter, the narrow relation that exists with the new control console and the mechanism is developed in the thired chapter, this relation from a point of view of an assembly of components is presents in fourth chapter, finally reaches and perspectives of mechanism forming part of project of the automation of reactor TRIGA MARK III, are present in the fifth chapter. (Author).

  17. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  18. 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

  19. Control of reactor coolant flow path during reactor decay heat removal

    International Nuclear Information System (INIS)

    Hunsbedt, A.N.

    1988-01-01

    This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

  20. The research reactors their contribution to the reactors physics

    International Nuclear Information System (INIS)

    Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

    2000-01-01

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  1. Technical review on irradiation tests and post-irradiation examinations in JMTR

    International Nuclear Information System (INIS)

    2017-07-01

    The Japan Materials Testing Reactor (JMTR) has been contributing to various R and D activities in the nuclear research such as the fundamental research of nuclear materials/ fuels, safety research and development of power reactors, radio isotope (RI) production since its beginning of the operation in 1968. Irradiation technologies and post irradiation examination (PIE) technologies are the important factors for irradiation test research. Moreover, these technologies induce the breakthrough in area of nuclear research. JMTR has been providing unique capabilities for the irradiation test research for about 40 years since 1968. In future, any needs for irradiation test research used irradiation test reactors will continue, such as R and D of generation 4 power reactors, fundamental research of materials/fuels, RI production. Now, decontamination and new research reactor construction are common issue in the world according to aging. This situation is the same in Japan. This report outlines irradiation and PIE technologies developed at JMTR in 40 years to contribute to the technology transfer and human resource development. We hope that this report will be used for the new research rector design as well as the irradiation test research and also used for the human resource development of nuclear engineers in future. (author)

  2. Joint European contribution to phase 5 of the BN600 hybrid reactor benchmark core analysis (European ERANOS formulaire for fast reactor core analysis)

    International Nuclear Information System (INIS)

    Rimpault, G.

    2004-01-01

    Hybrid UOX/MOX fueled core of the BN-600 reactor was endorsed as an international benchmark. BFS-2 critical facility was designed for full size simulation of core and shielding of large fast reactors (up tp 3000 MWe). Wide experimental programme including measurements of criticality, fission rates, rod worths, and SVRE was established. Four BFS-62 critical assemblies have been designed to study changes in BN-600 reactor physics-when moving to a hybrid MOX core. BFS-62-3A assembly is a full scale model of the BN-600 reactor hybrid core. it consists of three regions of UO 2 fuel, axial and radial fertile blankets, MOX fuel added in a ring between MC and OC zones, 120 deg sector of stainless steel reflector included within radial blanket. Joint European contribution to the Phase 5 benchmark analysis was performed by Serco Assurance Winfrith (UK) and CEA Cadarache (France). Analysis was carried out using Version 1.2 of the ERANOS code; and data system for advanced and fast reactor core applications. Nuclear data is based on the JEF2.2 nuclear data evaluation (including sodium). Results for Phase 5 of the BN-600 benchmark have been determined for criticality and SVRE in both diffusion and transport theory. Full details of the results are presented in a paper posted on the IAEA Business Collaborator website nad a brief summary is provided in this paper

  3. Reactor noise analysis of experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Ohtani, Hideji; Yamamoto, Hisashi

    1980-01-01

    As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

  4. Hot cell facilities for post irradiation examination

    International Nuclear Information System (INIS)

    Mishra, Prerna; Bhandekar, Anil; Pandit, K.M.; Dhotre, M.P.; Rath, B.N.; Nagaraju, P.; Dubey, J.S.; Mallik, G.K.; Singh, J.L.

    2017-01-01

    Reliable performance of nuclear fuels and critical core components has a large bearing on the economics of nuclear power and radiation safety of plant operating personnel. In view of this, Post Irradiation Examination (PIE) is periodically carried out on fuels and components to generate feedback information which is used by the designers, fabricators and the reactor operators to bring about changes for improved performance of the fuel and components. Examination of the fuel bundles has to be carried out inside hot cells due to their high radioactivity

  5. 15 años de Filosofía para la Paz. El lugar de la ética en la Investigación para la Paz

    Directory of Open Access Journals (Sweden)

    Ismael Cortés Gómez

    2014-05-01

    Full Text Available El propósito general de este artículo es explicar el modo en que la investigación ética se ha constituido epistemológica e institucionalmente como una disciplina científica, en el marco de la Cátedra UNESCO de Filosofía para la Paz. Este objetivo general se articula a partir de tres objetivos específicos: 1. Situar históricamente los principales modelos de análisis que han orientado la Investigación para la Paz en el siglo XX. 2. Explicar el origen del proyecto de Filosofía para la Paz en la década del noventa, en el contexto de la Europa democrática post-Socialista. 3. Analizar el diálogo que ha mantenido la Filosofía para la Paz con la Ética del Discurso y la Teoría Social del Reconocimiento.

  6. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Grover, S. Blaine

    2009-01-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  7. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  8. Cold neutron source conceptual designing for Tehran Research Reactor

    International Nuclear Information System (INIS)

    Khajvand, N.; Mirvakili, S.M.; Faghihi, F.

    2016-01-01

    Highlights: • Cold neutron source conceptual designing for Tehran research reactor is carried out. • Type and geometry of moderator and dimensions of cold neutron source are analyzed. • Liquid hydrogen with more ortho-concentration can be better option as moderator. - Abstract: A cold neutron source (CNS) conceptual designing for the Tehran Research Reactor (TRR) were carried out using MCNPX code. In this study, a horizontal beam tube of the core which has appropriate the highest thermal flux is selected and parametric analysis to choose the type and geometry of the moderator, and the required CNS dimensions for maximizing the cold neutron production was performed. In this design the moderator cell has a spherical annulus structure, and the cold neutron flux and its brightness are calculated together with the nuclear heat load of the CNS for a variety of materials including liquid hydrogen, liquid deuterium, and solid methane. Based on our study, liquid hydrogen with more ortho-concentration than para and solid methane are the best options.

  9. Injerto escalonado de calota para manejo de enoftalmos y distopia postraumática Tiered skull graft for the management of post-traumatic enophthalmos and dystopia

    Directory of Open Access Journals (Sweden)

    O.A. Vega Lagos

    2008-12-01

    Full Text Available Casi un 90%¹ de los traumatismos del macizo craneofacial involucran las órbitas, y un alto porcentaje de los mismos generan secuelas como son el enoftalmos y la distopia postraumáticos. Existen en la literatura numerosas referencias de técnicas quirúrgicas para corrección de dichas secuelas y diversos materiales para la reconstrucción orbitaria posterior al trauma, pero pocas de ellas se han encaminado a devolver la anatomía original al suelo orbitario, por tal razón se considera perentorio desarrollar una técnica sencilla y eficaz para la corrección de los defectos antes descritos. El presente artículo refiere la utilización de una nueva técnica quirúrgica para el manejo de las secuelas ya mencionadas usando un injerto escalonado de calota en el suelo de la órbita, dado que este último es una excelente opción para la colocación de injertos, debido a su baja tasa de reabsorción y mínimas reacciones adversas. Pacientes. Cuatro pacientes (1 mujer, 3 hombres con un promedio de edad de 23 años (rango entre 17 y 27 años con antecedente de trauma facial y compromiso de la órbita, que fueron tratados entre 2004 y 2007, en el Hospital Central de la Policía Nacional y Hospital El Tunal (Bogotá, Colombia, previa autorización mediante firma de consentimiento informado. En ellos se utilizó la técnica de injerto escalonado de calota para corrección de enoftalmos y distopia postraumática. Se describen resultados con un tiempo de seguimiento entre 11 y 42 semanas. Resultados. En todos los pacientes se observó disminución del continente orbitario y proyección simétrica del globo ocular en sentido antero-posterior y vertical, así como también mejoría completa de la diplopía, luego de la colocación del injerto escalonado de calota en el suelo orbitario. Conclusión. Se trata de una técnica sencilla, de bajo costo, mínima morbilidad, con resultados predecibles y satisfactorios a corto y medio plazo.Almost 90%¹ of

  10. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    Takahashi, Kazumi

    1998-01-01

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  11. Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Ott, Larry J.; Ellis, Ronald James; McDuffee, Joel Lee; Spellman, Donald J.; Bevard, Bruce Balkcom

    2009-01-01

    The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

  12. Method of fission product beta spectra measurements for predicting reactor anti-neutrino emission

    Energy Technology Data Exchange (ETDEWEB)

    Asner, D.M.; Burns, K.; Campbell, L.W.; Greenfield, B.; Kos, M.S., E-mail: markskos@gmail.com; Orrell, J.L.; Schram, M.; VanDevender, B.; Wood, L.S.; Wootan, D.W.

    2015-03-11

    The nuclear fission process that occurs in the core of nuclear reactors results in unstable, neutron-rich fission products that subsequently beta decay and emit electron antineutrinos. These reactor neutrinos have served neutrino physics research from the initial discovery of the neutrino to today's precision measurements of neutrino mixing angles. The prediction of the absolute flux and energy spectrum of the emitted reactor neutrinos hinges upon a series of seminal papers based on measurements performed in the 1970s and 1980s. The steadily improving reactor neutrino measurement techniques and recent reconsiderations of the agreement between the predicted and observed reactor neutrino flux motivates revisiting the underlying beta spectra measurements. A method is proposed to use an accelerator proton beam delivered to an engineered target to yield a neutron field tailored to reproduce the neutron energy spectrum present in the core of an operating nuclear reactor. Foils of the primary reactor fissionable isotopes placed in this tailored neutron flux will ultimately emit beta particles from the resultant fission products. Measurement of these beta particles in a time projection chamber with a perpendicular magnetic field provides a distinctive set of systematic considerations for comparison to the original seminal beta spectra measurements. Ancillary measurements such as gamma-ray emission and post-irradiation radiochemical analysis will further constrain the absolute normalization of beta emissions per fission. The requirements for unfolding the beta spectra measured with this method into a predicted reactor neutrino spectrum are explored.

  13. Design of the HMI for the operation of a nuclear research reactor; Diseno del HMI para la operacion de un reactor nuclear de investigacion

    Energy Technology Data Exchange (ETDEWEB)

    Bucio V, F. J.; Celis del Angel C, L.; Palacios H, J. C., E-mail: francisco.bucio@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2017-09-15

    The Instituto Nacional de Investigaciones Nucleares (ININ) participated in an international tender published by the Colombian Geological Service for the modernization of the Nuclear Reactor Control Console Ian-R1, the participating institutions were: General Atomics (USA), INVAP (Argentina) and ININ (Mexico). The proposal made by the ININ had an important characteristic, the independence of the manufacturer, since it was a project based on modular elements. One of the elements was the Human-Machine Interface (HMI), where the development was proposed under the Free Software (Gnu-GLP) scheme. Java was the programming language on which the HMI was developed to operate the nuclear reactor in Bogota, Colombia. The instrumentation that allows the interaction with the sensors and/or actuators is based on the use of PLC's (programmable logic controllers) with which the computers of the HMI communicate through a local network using the Mod bus protocol over Ethernet. (Author)

  14. Activación del topacio natural irradiado por neutrones en el núcleo del reactor RP-10

    OpenAIRE

    Gómez, J.; Parreño, Fernando; Lázaro, Gerardo; Vela, Mariano

    2003-01-01

    Se obtuvieron cristales de topacio activados al ser irradiados con neutrones dentro del núcleo del reactor RP-10. La activación depende del flujo de neutrones, por ello se desarrolló portamuestras (canes de irradiación) para absorber que son los causantes de la activación

  15. REMOCIÓN DE ARSÉNICO (V ASISTIDA POR OXIDACIÓN UV SOLAR EN UN FOTO-REACTOR TUBULAR DE SECCIÓN CIRCULAR

    Directory of Open Access Journals (Sweden)

    Ramiro Escalera Vásquez

    2010-01-01

    Full Text Available Se ha construido y caracterizado un foto-reactor tubular de sección circular para su aplicación al tratamiento de aguas subterráneas contaminadas con Arsénico, As(V, utilizando las técnica de la Remoción de Arsénico por Oxidación Solar (RAOS. El concentrador solar que posee una capacidad de radiación equivalente a 2,8 soles, fue construido reciclando materiales desechados: tubos de vidrio proveniente de lámparas de Ne y tubos de desagüe sanitario de 6” (PVC, recubiertos por láminas de aluminio. Pruebas simultáneas sin agitación,realizadas aplicando la radiación UV solar a aguas sintéticas, demostraron que la remoción de As(V en el foto-reactor es más rápida queen un tubo de vidrio sólo y en una botella PET de 2 litros, logrando remociones mayores al 98% en todos los casos. Los tiempos para la aparición de los flóculos de complejo Fe-citrato fueron de 40, 50 y 90 min respectivamente, para intensidades de radiación UVA integral (290-390 nm entre 50 y 70 Wm-2. Pruebas de irradiación seguidas de agitación controlada a 30-33 s-1 de gradiente de velocidad, demostraron que el foto-reactor acelera el proceso de formación de flóculos fácilmente sedimentables al cabo de 20-30 min de agitación. Los tiempos de irradiación óptimos para el foto-reactor, el tubo y la botella son de 15, 25 y 60 min, respectivamente. Pruebas en régimen de flujo continuo en un foto-reactor de aproximadamente 1 m2 de área, con un tiempo de residencia hidráulica (igual al tiempo de irradiación de 15 min, mostraron la formación inmediata de flóculos fácilmente sedimentables cuando se agitan a 33 s-1 durante 20-30 min, lográndose una remoción del 98,36% una concentración remanente de 16,5 mgL-1 de As(V en aguas decantadas. Esto significa que se pueden tratar aproximadamente 130 Lm-2 en una jornada de 6 horas de radiación UVA de 50-70 Wm-2 de intensidad.

  16. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  17. Diseño de biorreactores para producir bioetanol a partir de residuos de piel de patata

    OpenAIRE

    Pérez Madroñal, Rafael Ángel

    2018-01-01

    El objetivo principal de este proyecto es el dimensionamiento de los reactores de hidrólisis y fermentación de una planta de bioetanol para poder producir 27 millones de litros al año de bioetanol, teniendo como materia prima residuos de piel de patata. Para obtener el bioetanol, se recoge el almidón presente en los residuos de la piel de patata y se hacen pasar por dos reacciones, un proceso de hidrólisis enzimática y un proceso de fermentación. En ambos se usarán enzimas de l...

  18. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  19. IMPLEMENTACIÓN EN HARDWARE DE UN SUMADOR DE PUNTO FLOTANTE BASADO EN EL ESTÁNDAR IEEE 754-2008

    Directory of Open Access Journals (Sweden)

    Juan José Raygoza P.

    2009-01-01

    Full Text Available Este artículo presenta el diseño de un sumador de punto flotante descrito en lenguaje VHDL, basado en el estándar para Aritmética de Punto Flotante de IEEE (754¿-2008 para microprocesadores, del cual se utiliza el formato binario para precisión simple de 32 bits. El estándar define formatos para representar diferentes tipos de datos los cuales son: normal, subnormal, cero positivo, cero negativo, infinito positivo, infinito negativo y un no número (NaN. Muchas aplicaciones basadas en procesadores embebidos requieren la capacidad para realizar operaciones aritméticas de punto flotante, lo cual es fundamental para una mejor precisión y desempeño del sistema en el procesamiento de los datos. El sumador ha sido diseñado considerando los parámetros de velocidad, área utilizada dentro de la FPGA y consumo de potencia estimada; además el circuito ha sido sintetizado y simulado sobre las FPGAs Spartan®3 (3s200ft256-4, Virtex® II (2v1000fg256-4 y Virtex® 4 (4vfx12sf363-12 de la familia Xilinx®. El sumador ha sido diseñado por bloques de modo que podamos optimizar el proceso de cálculo por medio de las líneas de control, para que sólo la unidad indicada procese los datos. El circuito ha sido interconectado en un diagrama esquemático principal para la fácil incorporación de los bloques de control, entradas, salidas, cálculo simbólico y aritmético.

  20. FBR type reactor

    International Nuclear Information System (INIS)

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  1. RB Research nuclear reactor RB reactor, Annual report for 2000

    International Nuclear Information System (INIS)

    Milosevic, M.

    2000-12-01

    Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

  2. Procedures and techniques for the management of experimental fuels from research and test reactors. Proceedings of an advisory group meeting

    International Nuclear Information System (INIS)

    1999-04-01

    Almost all countries that have undertaken fuel development programs for power, research or military reactors have experimental and exotic fuels, either stored at the original research reactors where they have been tested or at some away-from-reactor storage facility. These spent fuel liabilities cannot follow the standard treatment recognized for modern power reactor fuels. They include experimental and exotic fuels ranging from liquids to coated spheres and in configurations ranging from full test assemblies to post irradiation examination specimens set in resin. This document contains an overview of the extent of the problem of managing experimental and exotic fuels from research and test reactors and an expert evaluation of the overall situation in countries which participated in the meeting

  3. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

    Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

  4. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  5. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

  6. New facility for post irradiation examination of neutron irradiated beryllium

    International Nuclear Information System (INIS)

    Ishitsuka, Etsuo; Kawamura, Hiroshi

    1995-01-01

    Beryllium is expected as a neutron multiplier and plasma facing materials in the fusion reactor, and the neutron irradiation data on properties of beryllium up to 800 degrees C need for the engineering design. The acquisition of data on the tritium behavior, swelling, thermal and mechanical properties are first priority in ITER design. Facility for the post irradiation examination of neutron irradiated beryllium was constructed in the hot laboratory of Japan Materials Testing Reactor to get the engineering design data mentioned above. This facility consist of the four glove boxes, dry air supplier, tritium monitoring and removal system, storage box of neutron irradiated samples. Beryllium handling are restricted by the amount of tritium;7.4 GBq/day and 60 Co;7.4 MBq/day

  7. Post-Irradiation Non-Destructive Analyses of the AFIP-7 Experiment

    Science.gov (United States)

    Williams, W. J.; Robinson, A. B.; Rabin, B. H.

    2017-12-01

    This article reports the results and interpretation of post-irradiation non-destructive examinations performed on four curved full-size fuel plates that comprise the AFIP-7 experiment. These fuel plates, having a U-10 wt.%Mo monolithic design, were irradiated under moderate operating conditions in the Advanced Test Reactor to assess fuel performance for geometries that are prototypic of research reactor fuel assemblies. Non-destructive examinations include visual examination, neutron radiography, profilometry, and precision gamma scanning. This article evaluates the qualitative and quantitative data taken for each plate, compares corresponding data sets, and presents the results of swelling analyses. These characterization results demonstrate that the fuel meets established irradiation performance requirements for mechanical integrity, geometric stability, and stable and predictable behavior.

  8. Subchannel analysis in nuclear reactors

    International Nuclear Information System (INIS)

    Ninokata, H.; Aritomi, M.

    1992-01-01

    This book contains 10 informative papers, presented at the International Seminar on Subchannel Analysis 1992 (ISSCA '92), organized by the Institute of Applied Energy, in collaboration with Atomic Energy Society of Japan, Tokyo Electric Power Company, Kansai Electric Power Company, Nuclear Power Engineering Corporation and the Japan Atomic Energy Research Institute, and held at the TIS-Green Forum, Tokyo, Japan, 30 October 1992. The seminar ISSCA '92 was intended to review the current state-of-the-arts of the method being applied to advanced nuclear reactors including Advanced BWRs, Advanced PWRs and LMRs, and to identify the problems to be solved, improvements to be made, and the needs of R and Ds that were required from the new fuel bundles design. The critical review was to focus on the performances of currently available subchannel analysis codes with regard to heat transfer and fluid flows in various types of nuclear reactor bundles under both steady-state and transient operating conditions, CHF, boiling transition (BT) or dryout behaviors and post BT. The behaviors of physical modeling and numerical methods in these extreme conditions were discussed and the methods critically evaluated in comparison with experiments. (author) (J.P.N.)

  9. The European fusion program and the role of the research reactors

    International Nuclear Information System (INIS)

    Laesser, R.; Andreani, R.; Diegele, E.

    2005-01-01

    The main objectives of the European long-term Fusion Technology Program are i) investigation of DEMO breeding blankets options, ii) development of low activation materials resistant to high neutron fluence, iii) construction of IFMIF for validation of DEMO materials, and iv) promotion of modelling efforts for the understanding of radiation damage. A large effort is required for the development and performance verification of the materials subjected to the intense neutron irradiation encountered in fusion reactors. In the absence of a strong 14.1 MeV neutron source fission materials research reactors are used. Elaborate in-pile and post-irradiation examinations are performed. In addition, the modelling effort is increased to predict the damage by a 'true' fusion spectrum in the future. Even assuming that a positive decision for IFMIF construction can be reached, the operation of a limited number of materials test reactors is needed to perform irradiation studies on large samples and for screening. (author)

  10. Para-Hermitian and para-quaternionic manifolds

    International Nuclear Information System (INIS)

    Ivanov, S.; Zamkovoy, S.

    2003-10-01

    A set of canonical para-Hermitian connections on an almost para-Hermitian manifold is defined. A Para-hermitian version of the Apostolov-Gauduchon generalization of the Goldberg-Sachs theorem in General Relativity is given. It is proved that the Nijenhuis tensor of a Nearly para-Kaehler manifolds is parallel with respect to the canonical connection. Salamon's twistor construction on quaternionic manifold is adapted to the para-quaternionic case. A locally conformally hyper-para-Kaehler (hypersymplectic) flat structure with parallel Lee form on the Kodaira-Thurston complex surfaces modeled on S 1 x SL (2, R)-tilde is constructed. Anti-self-dual locally conformally hyper-para-Kaehler (hypersymplectic) neutral metrics with non vanishing Weyl tensor are obtained on the Inoe surfaces. An example of anti-self-dual neutral metric which is not locally conformally hyper-para-Kaehler (hypersymplectic) is constructed. (author)

  11. Para-Hermitian and para-quaternionic manifolds

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, S [University of Sofia ' St. Kl. Ohridski' , Faculty of Mathematics and Informatics, Sofia (Bulgaria) and Abdus Salam International Centre for Theoretical Physics, Trieste (Italy); Zamkovoy, S [University of Sofia ' St. Kl. Ohridski' , Faculty of Mathematics and Informatics, Sofia (Bulgaria)

    2003-10-01

    A set of canonical para-Hermitian connections on an almost para-Hermitian manifold is defined. A Para-hermitian version of the Apostolov-Gauduchon generalization of the Goldberg-Sachs theorem in General Relativity is given. It is proved that the Nijenhuis tensor of a Nearly para-Kaehler manifolds is parallel with respect to the canonical connection. Salamon's twistor construction on quaternionic manifold is adapted to the para-quaternionic case. A locally conformally hyper-para-Kaehler (hypersymplectic) flat structure with parallel Lee form on the Kodaira-Thurston complex surfaces modeled on S{sup 1} x SL (2, R)-tilde is constructed. Anti-self-dual locally conformally hyper-para-Kaehler (hypersymplectic) neutral metrics with non vanishing Weyl tensor are obtained on the Inoe surfaces. An example of anti-self-dual neutral metric which is not locally conformally hyper-para-Kaehler (hypersymplectic) is constructed. (author)

  12. Shadow corrosion testing in the INCA facility in the Studsvik R2 reactor

    International Nuclear Information System (INIS)

    Nystrand, A.C.; Lassing, A.

    1999-01-01

    Shadow corrosion is a phenomenon which occurs when zirconium alloys are in contact with or in proximity to other metallic objects in a boiling water reactor environment (BWR, RBMK, SGHWR etc.). An enhanced corrosion occurs on the zirconium alloy with the appearance of a 'shadow' of the metallic object. The magnitude of the shadow corrosion can be significant, and is potentially limiting for the lifetime of certain zirconium alloy components in BWRs and other reactors with a similar water chemistry. In order to evaluate the suitability of the In-Core Autoclave (INCA) in the Studsvik R2 materials testing reactor as an experimental facility for studying shadow corrosion, a demonstration test has been performed. A number of test specimens consisting of Zircaloy-2 tubing in contact with Inconel were exposed in an oxidising water chemistry. Some of the specimens were placed within the reactor core and some above the core. The conclusion of this experiment after post irradiation examination is that it is possible to use the INCA facility in the Studsvik R2 reactor to develop a significant level of shadow corrosion after only 800 hours of irradiation. (author)

  13. Viability of inert matrix fuel in reducing plutonium amounts in reactors

    International Nuclear Information System (INIS)

    2006-08-01

    Reactors worldwide have produced more than 2000 tonnes of plutonium, contained in spent fuel or as separated forms through reprocessing. Disposition of fissile materials has become a primary concern of nuclear non-proliferation efforts. There is a significant interest in IAEA Member States to develop proliferation resistant nuclear fuel cycles for incineration of plutonium such as inert matrix fuels (IMFs). The present report summarises R and D work on inert matrix fuel for plutonium and (to a lesser extent) minor actinide stock-pile reduction, and discusses the possible strategies to include inert matrix fuel approaches to the nuclear fuel cycle. The publication reviews the status of potential IMF candidates and describes several identified candidate materials for both fast and thermal reactors: MgO, ZrO2, SiC, Zr alloy, SiAl, ZrN; some of these have undergone test irradiations and post-irradiation examination. Also discussed are modelling of IMF fuel performance and safety analysis. System studies have identified strategies for both implementation of IMF fuel as homogeneous or heterogeneous phases, as assemblies or core loadings and in existing reactors in the shorter term, as well as in new reactors in the longer term

  14. Evaluación del potencial acidogénico para producción de AGV de melaza de la industria azucarera como valorización de este subproducto

    Directory of Open Access Journals (Sweden)

    María Angélica Palomino

    2016-06-01

    Full Text Available Se evaluó el potencial acidogénico de la melaza de la industria azucarera en 4 diferentes OLR (6,02±4,33; 13,96±7,11; 15,81±4,83; 26,94±13,27kgDQO/m3.d en un reactor de fl ujo ascendente, con lodo granular. El sistema no contó con control de pH e inhibición de la fase metanogénica. El reactor operó en continuo durante 148 días. Para evaluar el potencial acidogénico se utilizó el grado de acidifi cación neto (GAn. Los resultados mostraron que durante las tres primeras OLR el %GAn (29,46 ± 13,01; 20,23 ± 13,67; 24,63 ± 19,49 se mantuvo sin diferencias signifi cativas, pero para la mayor OLR el %GAn disminuyó a la tercera parte (10,21 ± 7,14, mientras la concentración de AGV fue la mayor para esta fase (2644,89mgDQO/L, además se avaluó el balance de DQO para cada una de las fases, donde el % de AGV en el efl uente representó el porcentaje orgánico fermentable rápidamente en el efl uente, estos valores indican que con una recirculación interna se podría mejorar el %GAn u obtener otra serie de productos de base biológica para el aprovechamiento de este residuo. En este artículo se utilizó un reactor de fl ujo ascendente como alternativa a los estudiados (CSRT y batch presentando diferentes resultados

  15. Accelerator-driven transmutation reactor analysis code system (ATRAS)

    Energy Technology Data Exchange (ETDEWEB)

    Sasa, Toshinobu; Tsujimoto, Kazufumi; Takizuka, Takakazu; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    JAERI is proceeding a design study of the hybrid type minor actinide transmutation system which mainly consist of an intense proton accelerator and a fast subcritical core. Neutronics and burnup characteristics of the accelerator-driven system is important from a view point of the maintenance of subcriticality and energy balance during the system operation. To determine those characteristics accurately, it is necessary to involve reactions at high-energy region, which are not treated on ordinary reactor analysis codes. The authors developed a code system named ATRAS to analyze the neutronics and burnup characteristics of accelerator-driven subcritical reactor systems. ATRAS has a function of burnup analysis taking account of the effect of spallation neutron source. ATRAS consists of a spallation analysis code, a neutron transport codes and a burnup analysis code. Utility programs for fuel exchange, pre-processing and post-processing are also incorporated. (author)

  16. Nuclear reactors. Introduction

    International Nuclear Information System (INIS)

    Boiron, P.

    1997-01-01

    This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

  17. Nuclear reactor, reactor core thereof, and device for constituting the reactor

    International Nuclear Information System (INIS)

    Takiyama, Masashi.

    1994-01-01

    A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

  18. Competencias para el desarrollo sostenible: las capacidades, actitudes y valores meta de la educación en el marco de la Agenda global post-2015

    Directory of Open Access Journals (Sweden)

    María Ángeles Murga-Menoyo

    2015-07-01

    Full Text Available Este artículo se focaliza en la formación de las competencias y capacidades que precisan las personas para construir sociedades caracterizadas por la sostenibilidad de su desarrollo. Propone una matriz competencial básica construida a partir de las cuatro competencias que la Unesco considera clave para afrontar este reto: análisis crítico, reflexión sistémica, toma de decisión colaborativa y sentido de responsabilidad hacia las generaciones presentes y futuras. En el marco de los procesos de enseñanza-aprendizaje, la autora entiende cada una de ellas como resultado de una pluralidad de factores, a su vez, compuestos por distintas capacidades que, como fruto de los procesos formativos, los estudiantes pueden manifestar en comportamientos observables (logros de aprendizaje. La matriz se completa con cuatro rúbricas que recogen indicadores (evidencias significativos en el desempeño de la correspondiente competencia. Estas rúbricas se conciben como un instrumento al servicio del proceso formativo, y, en especial, del aprendizaje autorregulado. La propuesta, abierta y versátil, puede ser adaptada a diferentes contextos y circunstancias. Pretende contribuir a una reorientación de la práctica docente hacia el desarrollo sostenible, que pueda ser asumida por el profesorado de todos los niveles educativos, tanto del sistema escolar como de la formación profesional y la universitaria.Cómo referenciar este artículoMurga-Menoyo, M. A. (2015. Competencias para el desarrollo sostenible: las capacidades, actitudes y valores meta de la educación en el marco de la Agenda global post-2015. Foro de Educación, 13(19, 55-83. doi: http://dx.doi.org/10.14516/fde.2015.013.019.004

  19. Life management for a non replaceable structure: the reactor building

    International Nuclear Information System (INIS)

    Torres, V.; Francia, L.

    1998-01-01

    Phase 1 of UNESA N.P.P. Lifetime Management Project identified and ranked important components, relative to plant life management. The list showed the Reactor Containment Structure in the third position, and thirteen concrete structures were among the list top twenty. Since the Reactor Containment Building, together with the Reactor Vessel, is the only non-replaceable plant component, and has a big impact on the plant technical life, there is an increasing interest on understanding its behavior to maintain structural integrity. This paper presents: a) IAEA (International Atomic Energy Agency) Coordinated Research Program experiences and studies. Under this Program, international experts address the most frequent degradation mechanisms affecting the containment building. b) IAEA Aging Management Program adapted to our plants. The paper addresses the aging mechanisms affecting the concrete structures, reinforcing steel and prestress systems as well as the aging management programs and the mitigation and control methods. Finally, this paper presents a new module called STRUCTURES, included in phase 2 of the above mentioned project, which will monitor and document the different aging mechanisms and management programs described in item b) regarding the Reactor Containment Building (concrete liner, post stressing system, anchor elements). This module will also support the Maintenance Rule related practices. (Author)

  20. A high resolution pneumatic stepping actuator for harsh reactor environments

    Science.gov (United States)

    Tippetts, Thomas B.; Evans, Paul S.; Riffle, George K.

    1993-01-01

    A reactivity control actuator for a high-power density nuclear propulsion reactor must be installed in close proximity to the reactor core. The energy input from radiation to the actuator structure could exceed hundreds of W/cc unless low-cross section, low-absorptivity materials are chosen. Also, for post-test handling and subsequent storage, materials should not be used that are activated into long half-life isotopes. Pneumatic actuators can be constructed from various reactor-compatible materials, but conventional pneumatic piston actuators generally lack the stiffness required for high resolution reactivity control unless electrical position sensors and compensated electronic control systems are used. To overcome these limitations, a pneumatic actuator is under development that positions an output shaft in response to a series of pneumatic pulses, comprising a pneumatic analog of an electrical stepping motor. The pneumatic pulses are generated remotely, beyond the strong radiation environment, and transmitted to the actuator through tubing. The mechanically simple actuator uses a nutating gear harmonic drive to convert motion of small pistons directly to high-resolution angular motion of the output shaft. The digital nature of this actuator is suitable for various reactor control algorithms but is especially compatible with the three bean salad algorithm discussed by Ball et al. (1991).

  1. Slurry reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

    1979-08-01

    Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

  2. Optimal reactor strategy for commercializing fast breeder reactors

    International Nuclear Information System (INIS)

    Yamaji, Kenji; Nagano, Koji

    1988-01-01

    In this paper, a fuel cycle optimization model developed for analyzing the condition of selecting fast breeder reactors in the optimal reactor strategy is described. By dividing the period of planning, 1966-2055, into nine ten-year periods, the model was formulated as a compact linear programming model. With the model, the best mix of reactor types as well as the optimal timing of reprocessing spent fuel from LWRs to minimize the total cost were found. The results of the analysis are summarized as follows. Fast breeder reactors could be introduced in the optimal strategy when they can economically compete with LWRs with 30 year storage of spent fuel. In order that fast breeder reactors monopolize the new reactor market after the achievement of their technical availability, their capital cost should be less than 0.9 times as much as that of LWRs. When a certain amount of reprocessing commitment is assumed, the condition of employing fast breeder reactors in the optimal strategy is mitigated. In the optimal strategy, reprocessing is done just to meet plutonium demand, and the storage of spent fuel is selected to adjust the mismatch of plutonium production and utilization. The price hike of uranium ore facilitates the commercial adoption of fast breeder reactors. (Kako, I.)

  3. Computerized reactor monitor and control for nuclear reactors

    International Nuclear Information System (INIS)

    Buerger, L.

    1982-01-01

    The analysis of a computerized process control system developed by Transelektro-KFKI-Videoton (Hangary) for a twenty-year-old research reactor in Budapest and or a new one in Tajura (Libya) is given. The paper describes the computer hardware (R-10) and the implemented software (PROCESS-24K) as well as their applications at nuclear reactors. The computer program provides for man-machine communication, data acquisition and processing, trend and alarm analysis, the control of the reactor power, reactor physical calculations and additional operational functions. The reliability and the possible further development of the computerized systems which are suitable for application at reactors of different design are also discussed. (Sz.J.)

  4. Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

  5. Síntesis de membranas cerámicas para la regeneración de baños de cromado agotados

    Directory of Open Access Journals (Sweden)

    Sánchez, E.

    2005-12-01

    Full Text Available Ceramic membranes intended for use as compartment separators in electrochemical reactors used for recycling spent chromium plating baths have been synthesised. Two variables of the membrane synthesis process have been studied (pressing pressure and organic matter addition, to enable designing prototypes with the appropriate characteristics to act as separators, at a low manufacturing cost.

    Se han sintetizado membranas cerámicas destinadas a la función de separadores entre los compartimentos de un reactor electroquímico, cuya aplicación es el reciclado de baños de cromado usados. Se han estudiado dos variables del proceso de síntesis de las membranas (presión de prensado y adición de materia orgánica, para conseguir prototipos con las características adecuadas para realizar la función de separador, y a la vez conseguir un coste de fabricación reducido.

  6. Albanian Foreign Policy in the Post-Comunist Era

    Directory of Open Access Journals (Sweden)

    Abdurrahim F. Aydin

    2011-05-01

    Full Text Available Estando bajo un régimen totalitario durante el periodo comunista, la política exterior albanesa sacrificó sus objetivos políticos y sus intereses nacionales. Con el colapso del Comunismo, el liderazgo albanés tiene la encomiable tarea de relacionar correctamente los objetivos reales y los intereses genuinos de los albaneses en Albania, Serbia, Kosovo, KYROM, Montenegro y Grecia. Los propósitos de este artículo sn enfatizar la política exterior albanesa actual (post-comunista y mostrar qué falta en la política exterior de Albania para encajar sus intereses nacionales en sus continuos esfuerzos para establecer una exitosa europeización y democratización de su cultura.

  7. Control of reactor coolant flow path during reactor decay heat removal

    Science.gov (United States)

    Hunsbedt, Anstein N.

    1988-01-01

    An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

  8. Nupec thermal hydraulic test to evaluate post-DNB characteristics for PWR fuel assemblies (1. general test plan and results)

    International Nuclear Information System (INIS)

    Norio, Kono; Kenji, Murai; Kaichiro, Misima; Takayuki, Suemura; Yoshiei, Akiyama; Keiichi, Hori

    2001-01-01

    In the present thermal hydraulic design of Pressurized Water Reactor (PWR), a departure from nucleate boiling (DNB) under anticipated transient conditions is not allowed. However, it is recognized that the DNB dose not cause a fuel rod failure immediately, and a suitable reactor trip can prevent the core from severe damages. If the fuel rod temperature under the post-DNB conditions can be accurately evaluated, the potentially existing margin in the present design method will be quantitatively assessed. To establish the heat transfer evaluation method on post-DNB event for PWR thermal hydraulic design, Nuclear Power Engineering Corporation (NUPEC) started a program, NUPEC Thermal Hydraulic Test to Evaluate Post-DNB Characteristics for PWR Fuel Assemblies (NUPEC-TH-P), in 1995 (hereinafter the year means fiscal year) under the sponsorship of Ministry of Economy, Trade and industry (METI). This program is now under going until 2001. This paper is to show the overall plan and the status of NUPEC-TH-P. (authors)

  9. Simulation of the concentration of SO{sub 2} issued by major stationary sources in 2003 in northwestern Chiapas and central Tabasco, Mexico; Simulacion de la concentracion de SO{sub 2} emitido por fuentes fijas mayores durante 2003 en el noroeste de Chiapas y centro de Tabasco, Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Valdes Manzanilla, Arturo [Division Academica de Ciencias Biologicas, UJAT, Villahermosa, Tabasco (Mexico)]. E-mail: avmanzanilla@hotmail.com; Fernandez Garcia, German [Universidad Autonoma de Guadalajara Campus Tabasco, Villahermosa, Tabasco (Mexico); Ramos Herrera, Sergio [Division Academica de Ciencias Biologicas, UJAT, Villahermosa, Tabasco (Mexico); Bautista Margulis, Raul G. [Division Academica de Ciencias Biologicas, UJAT, Villahermosa, Tabasco (Mexico)

    2008-05-15

    The CALPUFF dispersion model was used to simulate the spatial distribution of the SO{sub 2} emitted by Petroleos Mexicanos (PEMEX) gas-processing complexes in Cactus, NW Chiapas, Nuevo PEMEX and Ciudad PEMEX, in central Tabasco, during 2003. It was found that the zone with the highest concentrations of SO{sub 2} is located SW from the PEMEX facilities. The zone, where the World Health Organization's norm for SO{sub 2} is surpassed (20 {mu}g/m{sup 3}), includes the city of Reforma, Chiapas and Ciudad PEMEX, Tabasco. [Spanish] El modelo de dispersion CALPUFF fue usado para simular la distribucion espacial de la concentracion de SO{sub 2} emitido por los complejos procesadores de gas de Petroleos Mexicanos (PEMEX) en Cactus, en el noroeste de Chiapas, Nuevo PEMEX y Ciudad PEMEX, en el centro de Tabasco, durante 2003. Se encontro que la zona con mayor concentracion de SO{sub 2} esta localizada al suroeste de las instalaciones de la empresa petrolera. La zona, donde la norma de la Organizacion Mundial de la Salud para el SO{sub 2} es sobrepasada (20 {mu}g/m{sup 3}), incluye la ciudad de Reforma, Chiapas y Ciudad PEMEX, Tabasco.

  10. Non-Destructive Testing in Reactor Pressure-Vessel Fabrication; Essais non Destructifs dans la Fabrication des Caissons Etanches de Reacteurs; Nedestruktivnoe ispytanie pri izgotovlenii reaktornykh bakov vysokogo davleniya; Ensayo no Destructivo Durante la Fabricacion de Recipientes de Presion para Reactores

    Energy Technology Data Exchange (ETDEWEB)

    McGonnagle, W. J. [Fluids Dynamics Research, Iit Research Institute, Chicago, IL (United States)

    1965-09-15

    applicables. Il suggere des criteres, a la fois realistes et satisfaisants, d'acceptation et de rejet. Il expose les grandes lignes d'une procedure qui permettra au personnel charge des essais non destructifs d'accomplir sa tache de maniere appropriee au stade opportun du cycle de fabrication. Il etudie les rapports entre le groupe charge des essais non destructifs et les autres groupes de personnel intervenant dans la fabrication du caisson. (author) [Spanish] El presente trabajo tiene como finalidad esbozar brevemente un programa de control de calidad aplicado en el proyecto y construccion de un recipiente de presion para reactor, capaz de satisfacer todas las exigencias nucleares y de seguridad; asimismo se propone poner de manifiesto el papel y la importancia de los ensayos no destructivos en el logro de ese objetivo. Las fallas observadas en materiales, componentes y conjuntos de elementos, ponen de manifiesto que las actuales tecnicas de fabricacion no bastan por sf solas para garantizar en todos los casos la seguridad de servicio de los componentes criticos. Aun empleando los mejores procesos, asf como tambien metodos y tecnicas sometidas a controles apropiados, aparecen fallas y heterogeneidades. Por lo tanto, se requiere un programa adecuado y correctamente integrado de ensayos no destructivos, a fin de lograr el nivel de calidad imprescindible para el recipiente de presion de todo reactor nuclear. Los principales metodos no destructivos aplicados por los fabricantes de recipientes de presion para reactores son: inspeccion visual, radiografia y gammagraffa, ensayo ultrasonico, y empleo de particulas magneticas y de Ifquidos penetrantes. El programa de ensayos no destructivos incluye la inspeccion del material en forma de chapas, piezas forjadas, piezas coladas, revestimientos y soldaduras. Se analizan en este trabajo los problemas particulares con que tropieza el ensayo no destructivo aplicado a recipientes de presion para reactores nucleares. Se exponen y discuten

  11. Fuel development for reactors of new generation in Ukraine

    International Nuclear Information System (INIS)

    Odeychuk, N.P.

    2006-01-01

    Full text: On the background of critical situation in traditional power engineering due to deficiency of organic fuel, physical and moral ageing of the of thermal power stations equipment and their harmful influence on the ecology of environment, the nuclear engineering works stably enough and, by keeping all safety measures, is the most non-polluting energy source. In Ukraine the atomic engineering became one of main sources of energy production and is the important factor of guarantee the power engineering independence of the state. The main center on development of the components of nuclear reactors active zones is the National scientific center K harkov institute of Physics and Technology . The significant place in institutes' investigations was occupied with works on creation the constructional materials and nuclear fuel for heavy water reactors E-circumflexS-150, OR-1000, OR-2000, light water reactors WWER-1000 and RBMK-1500, high-temperature gas cooled reactors ABTU and HTGR, gas reactors on fast neutrons BGR and BRGD, and also the reactor - converter ROMASHKA and other special reactors of special assignment. Radiation tests and post-irradiation research confirm intended material-study, technological and design decisions and fuel elements capacity work on the whole. Nevertheless, by the present conditions, it is necessary to pay special attention to development of the new, safe guaranteed nuclear energy sources. In Ukraine proceed works on research and development of new safe nuclear reactors: basing the underground nuclear thermal power stations; development the reactors with managed chain reaction of nucleus division in an active zone with the help of an external source of neutrons; power thermonuclear installations; high-temperature helium reactors which are especially actual now from the point of view of the hydrogen production; the advanced pressure water reactors, heavy water reactors. In the paper also discussed the state of works in Ukraine on fuel

  12. Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

    International Nuclear Information System (INIS)

    Hauptman, H.M.; Petro, J.N.; Jacobi, O.

    1995-04-01

    This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

  13. Reactor core for FBR type reactor

    International Nuclear Information System (INIS)

    Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

    1996-01-01

    In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

  14. Space-time reactor kinetics for heterogeneous reactor structure

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1969-11-15

    An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods.

  15. La agenda post 2015: desafío de resiliencia que reorienta la cooperación global

    Directory of Open Access Journals (Sweden)

    Erli Margarita Marín Aranguren

    2013-11-01

    Full Text Available Los Objetivos de Desarrollo del Milenio (ODM, los cuales marcaron un hito en términos de la cooperación al desarrollo al haber logrado que el mundo se pusiera de acuerdo en cómo afrontar los grandes problemas actuales, están a punto de expirar. Por eso, en diferentes escenarios se avanza en propuestas para construir la agenda post 2015, la cual determinará un nuevo acuerdo internacional para adelantar la cooperación hasta 2030. En este artículo se pone en blanco y negro dicho proceso que ha tenido lugar en varios escenarios, con la participación de diferentes actores. De esta manera, se pretende mostrar una visión de conjunto a los lectores hispanohablantes. Primero se describen los alcances y las lecciones que dejó el proceso de los ODM, y de allí analiza cuánto se han aplicado los aprendizajes en los distintos escenarios donde se han elaborado propuestas para la nueva agenda. Finalmente, se señalan algunos retos para la construcción de la agenda post 2015 y se plantean nuevos interrogantes.

  16. Determination of hydrazine in third loops of China experimental fast reactor by spectrophotometry

    International Nuclear Information System (INIS)

    Huang Wenjie; Wang Mi; Gao Yaopeng; Xie Chun; Yu Xiaochen

    2013-01-01

    The method for the determination of hydrazine by Uv-vis spectrophotometer was proposed. The coloration conditions and instrument parameters were also optimized. In HCl, hydrazine formed a yellow azine with para-dimethyl aminobenzaldehyde ((CH 3 ) 2 NC 6 H 4 CHO), and then determined by spectrophotometer. The complex's maximum absorption was exhibited at 458 nm. The coloration effect was excellent in conditions of 1% HCl, 10 mL para-dimethyl aminobenzaldehyde and 10 minutes' developing time. A good liner relationship was obtained in the range of 5∼200 μg/L, and the recovery was (101.1±1.9)%. This method was used in the third loop of China experimental fast reactor with satisfactory results. (authors)

  17. Nuclear Reactor RA Safety Report, Vol. 4, Reactor

    International Nuclear Information System (INIS)

    1986-11-01

    RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

  18. Material Science Activities for Fusion Reactors in Kazakhstan

    International Nuclear Information System (INIS)

    Tazhibayeva, I.; Kenzhin, E.; Kulsartov, T.; Shestakov, V.; Chikhray, Y.; Azizov, E.; Filatov, O.; Chernov, V.M.

    2007-01-01

    Full text of publication follows: Paper contains results of fusion material testing national program and results of activities on creation of material testing spherical tokamak. Hydrogen isotope behavior (diffusion, permeation, and accumulation) in the components of the first wall and divertor was studied taking into account temperature, pressure, and reactor irradiation. There were carried out out-of-pile and in-pile (reactors IVG-IM, WWRK, RA) studies of beryllium of various grades (TV-56, TShG-56, DV-56, TGP-56, TIP-56), graphites (RG-T, MPG-8, FP 479, R 4340), molybdenum, tungsten, steels (Cr18Ni10Ti, Cr16Ni15, MANET, F82H), alloys V-(4-6)Cr-( 4-5)Ti, Cu+1%Cr+0.1%Zr, and double Be/Cu and triple Be/Cu/steel structures. Tritium permeability from eutectic Pb+17%Li through steels Cr18Ni10Ti, Cr16Ni15, MANET, and F82H were studied taking into account protective coating effects. The tritium production rate was experimentally assessed during in-pile and post-reactor experiments. There were carried out radiation tests of ceramic Li 2 TiO 3 (96% enrichment by Li-6) with in-situ registration of released tritium and following post-irradiation material tests of irradiated samples. Verification of computer codes for simulation of accidents related to LOCA in ITER reactor was carried out. Codes' verification was carried out for a mockup of first wall in a form of three-layer cylinder of beryllium, bronze (Cu-Cr-Zr) and stainless steel. At present Kazakhstan Tokamak for Material testing (tokamak KTM) is created in National Nuclear Center of Republic of Kazakhstan in cooperation with Russian Federation organizations (start-up is scheduled on 2008). Tokamak KTM allows for expansion and specification of the studies and tests of materials, protection options of first wall, receiving divertor tiles and divertor components, methods for load reduction at divertor, and various options of heat/power removal, fast evacuation of divertor volume and development of the techniques for

  19. Feasible reactor power cutback logic development for an integral reactor

    International Nuclear Information System (INIS)

    Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

    2013-01-01

    Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

  20. Post Irradiation Mechanical Behaviour of Three EUROFER Joints

    International Nuclear Information System (INIS)

    Lucon, E.; Leenaers, A.; Vandermeulen, W.

    2006-01-01

    The post-irradiation mechanical properties of three EUROFER joints (two diffusion joints and one TIG weld) have been characterized after irradiation to 1.8 dpa at 300 degrees Celsius in the BR-2 reactor. Tensile, KLST impact and fracture toughness tests have been performed. Based on the results obtained and on the comparison with data from EUROFER base material irradiated under similar conditions, the post-irradiation mechanical behaviour of both diffusion joints (laboratory and mock-up) appears similar to that of the base material. The properties of the TIG joint are affected by the lack of a post-weld heat treatment, which causes the material from the upper part of the weld to be significantly worse than that of the lower region. Thus, specimens from the upper layer exhibit extremely pronounced hardening and embrittlement caused by irradiation. The samples extracted from the lower layer show much better resistance to neutron exposure, although their measured properties do not match those of the diffusion joints. The results presented demonstrate that diffusion joining can be a very promising technique.

  1. RA Reactor

    International Nuclear Information System (INIS)

    1978-02-01

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

  2. The effect of harvesting on biomass production and nutrient removal in phototrophic biofilm reactors for effluent polishing

    NARCIS (Netherlands)

    Boelee, N.C.; Janssen, M.; Temmink, H.; Taparaviciute, L.; Khiewwijit, R.; Janoska, A.; Buisman, C.J.N.; Wijffels, R.H.

    2014-01-01

    An increasing number of wastewater treatment plants require post-treatment to remove residual nitrogen and phosphorus. This study investigated various harvesting regimes that would achieve consistent low effluent concentrations of nitrogen and phosphorus in a phototrophic biofilm reactor.

  3. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  4. Cleanup and decommissioning of a nuclear reactor after a severe accident

    International Nuclear Information System (INIS)

    1992-01-01

    Although the development of commercial nuclear power plants has in general been associated with an excellent record of nuclear safety, the possibility of a severe accident resulting in major fuel and core damage cannot be excluded and such accidents have in fact already occurred. For over a decade, IAEA publications have provided technical guidance and recommendations for post-accident planning to be considered by appropriate authorities. Guidance and recommendations have recently been published on the management of damaged nuclear fuel, sealing of the reactor building and related safety and performance assessment aspects. The present technical report on the cleanup and decommissioning of reactors which have undergone a severe accident represents a further publication in the series. Refs, figs and tabs.

  5. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

    The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

  6. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  7. Independent Confirmatory Survey Report for the University of Arizona Nuclear Reactor Laboratory, Tucson, Arizona DCN:2051-SR-01-0

    International Nuclear Information System (INIS)

    Altic, Nick A.

    2011-01-01

    The University of Arizona (University) research reactor is a TRIGA swimming pool type reactor designed by General Atomics and constructed at the University in 1958. The reactor first went into operation in December of 1958 under U.S. Nuclear Regulatory Commission (NRC) license R-52 until final shut down on May 18, 2010. Initial site characterization activities were conducted in February 2009 during ongoing reactor operations to assess the radiological status of the Nuclear Reactor Laboratory (NRL) excluding the reactor tank, associated components, and operating systems. Additional post-shutdown characterization activities were performed to complete characterization activities as well as verify assumptions made in the Decommissioning Plan (DP) that were based on a separate activation analysis (ESI 2009 and WMG 2009). Final status survey (FSS) activities began shortly after the issuance of the FSS plan in May 2011. The contractor completed measurement and sampling activities during the week of August 29, 2011.

  8. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    Science.gov (United States)

    Hill, P.R.

    1994-12-27

    A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

  9. Swelling Estimation of Multi-wire U-Mo Monolithic Fuel for HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon-Sang; Ryu, Ho-Jin; Park, Jong-Man; Oh, Jong-Myeong; Kim, Chang-Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-10-15

    In order to use low-enriched uranium (LEU) instead of highly enriched uranium (HEU) for high performance research reactors, the reduced enrichment for research and test reactors (RERTR) program is developing high uranium density fuel such as U-Mo/Al dispersion fuel. U-Mo alloys have an excellent irradiation performance when compared to other uranium alloys or compounds. But the results from the post-irradiation examination of the U-Mo/Al dispersion fuels indicate that an interaction between the U-Mo alloy fuel and the Al matrix phases occurs readily during an irradiation and it is sensitively dependent on the temperature. In order to lessen these severe interactions, a concept of a multi-wire type fuel was proposed. The fuel configuration is that three to six U-Mo fuel wires (1.5 mm - 2 mm in diameter) are symmetrically arranged at the periphery side in the Al matrix. In this study temperature calculations and a swelling estimation of a multi-wire monolithic fuel were carried out. Also the results of a post irradiation analysis of this fuel will be introduced.

  10. Final project report, TA-35 Los Alamos Power Reactor Experiment No. II (LAPRE II) decommissioning project

    International Nuclear Information System (INIS)

    Montoya, G.M.

    1992-01-01

    This final report addresses the decommissioning of the LAPRE II Reactor, safety enclosure, fuel reservoir tanks, emergency fuel recovery system, primary pump pit, secondary loop, associated piping, and the post-remediation activities. Post-remedial action measurements are also included. The cost of the project, including Phase I assessment and Phase II remediation was approximately $496K. The decommissioning operation produced 533 m 3 of low-level solid radioactive waste and 5 m 3 of mixed waste

  11. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

  12. CO_2 capture with solid sorbent: CFD model of an innovative reactor concept

    International Nuclear Information System (INIS)

    Barelli, L.; Bidini, G.; Gallorini, F.

    2016-01-01

    Highlights: • A new reactor solution based on rotating fixed beds was presented. • The preliminary design of the reactor was approached. • A CFD model of the reactor, including CO_2 capture kinetic, was developed. • The CFD model is validated with experimental results. • Sorbent exploitation increasing is possible thanks to the new reactor. - Abstract: In future decarbonization scenarios, CCS with particular reference to post-combustion technologies will be an important option also for energy intensive industries. Nevertheless, today CCS systems are rarely installed due to high energy and cost penalties of current technology based on chemical scrubbing with amine solvent. Therefore, innovative solutions based on new/optimized solvents, sorbents, membranes and new process designs, are R&D priorities. Regarding the CO_2 capture through solid sorbents, a new reactor solution based on rotating fixed beds is presented in this paper. In order to design the innovative system, a suitable CFD model was developed considering also the kinetic capture process. The model was validated with experimental results obtained by the authors in previous research activities, showing a potential reduction of energy penalties respect to current technologies. In the future, the model will be used to identify the control logic of the innovative reactor in order to verify improvements in terms of sorbent exploitation and reduction of system energy consumption.

  13. Resistencia de dientes restaurados con postes prefabricados ante cargas de máxima intercuspidación, masticación y bruxismo Resistance of teeth restored with prefabricated posts to maximum intercuspidation loads, mastication and bruxism

    Directory of Open Access Journals (Sweden)

    Santiago Correa Vélez

    2013-03-01

    Full Text Available Objetivo: determinar por el método de los elementos finitos la resistencia de dientes restaurados con postes prefabricados ante cargas estáticas de máxima intercuspidación y cargas cíclicas de masticación y bruxismo y analizar el efecto de la pérdida periodontal en la resistencia de las restauraciones. Métodos: se realizó una investigación in vitro mediante el método de los elementos finitos de dientes con pérdida periodontal, rehabilitados con postes prefabricados en fibra de vidrio, carbono y titanio. Los dientes fueron reconstruidos a partir de imágenes tomográficas de un paciente periodontalmente sano. Resultados: se muestra que ante cargas estáticas las rehabilitaciones no presentan tendencia a la falla, independientemente del material del poste o del grado de pérdida periodontal. En el caso de bruxismo y pérdida periodontal de 4 mm, la dentina presenta una durabilidad de 60 000 ciclos independiente del material del poste. Para cargas de masticación y periodonto sano, la falla en la dentina ocurre a los 100 000 ciclos con poste en titanio, 200 000 ciclos con poste en fibra de carbono y 1 100 000 ciclos con poste en fibra de vidrio. Para una pérdida periodontal de 2 mm la durabilidad de la dentina se reduce a 4 000 ciclos con poste en titanio, 5 000 ciclos con poste en fibra de carbono y 7 000 ciclos con poste en fibra de vidrio. Para pérdida periodontal de 4 mm, la durabilidad de la dentina se estima en 1 000 ciclos, independientemente del material del poste utilizado. Conclusiones: ante carga estática de máxima intercuspidación las rehabilitaciones con postes prefabricados en fibra de vidrio, carbono y titanio no presentan tendencia a la falla, independientemente del grado de pérdida periodontal. Ante cargas cíclicas, los postes prefabricados presentan una vida útil infinita, y es la dentina la estructura más afectada ante dichos eventos.Objective: using the finite element method, determine the resistance of

  14. TRACG post-test analysis of panthers prototype tests of SBWR passive containment condenser

    International Nuclear Information System (INIS)

    Fitch, J.R.; Billig, P.F.; Abdollahian, D.; Masoni, P.

    1997-01-01

    As part of the validation effort for application of the TRACG code to the Simplified Boiling Water Reactor (SBWR), calculations have been performed for the various test facilities which are part of the SBWR design and technology certification program. These calculations include post-test calculations for tests in the PANTHERS Passive Containment Condenser (PCC) test program. Sixteen tests from the PANTHERS/PCC test matrix were selected for post-test analysis. This set includes three steady-state pure-steam tests, nine steady-state steam-air tests, and four transient tests. The purpose of this paper is to present and discuss the results of the post-test analysis. The author includes a brief description of the PANTHERS/PCC test facility and test matrix, a description of the PANTHERS/PCC post-test TRACG model and the manner in which the various types of tests in the post-test evaluation were simulated, and a presentation of the results of the TRACG simulation

  15. Strategic planning for research reactors. Guidance for reactor managers

    International Nuclear Information System (INIS)

    2001-04-01

    The purpose of this publication is to provide guidance on how to develop a strategic plan for a research reactor. The IAEA is convinced of the need for research reactors to have strategic plans and is issuing a series of publications to help owners and operators in this regard. One of these covers the applications of research reactors. That report brings together all of the current uses of research reactors and enables a reactor owner or operator to evaluate which applications might be possible with a particular facility. An analysis of research reactor capabilities is an early phase in the strategic planning process. The current document provides the rationale for a strategic plan, outlines the methodology of developing such a plan and then gives a model that may be followed. While there are many purposes for research reactor strategic plans, this report emphasizes the use of strategic planning in order to increase utilization. A number of examples are given in order to clearly illustrate this function

  16. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

  17. Spectrographic determination of metallic impurities in organic coolants for nuclear reactors; Determinacion espectrografica de impurezas metalicas en refrigerantes organicos para reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Martin Munoz, M; Alvarez Gonzalez, F

    1969-07-01

    A spectrochemical method for determining metallic impurities in organic coolants for nuclear reactors is given. The organic matter in solid samples is eliminated by controlled distillation and dry ashing in the presence of magnesium oxide as carrier. Liquid, samples are vacuum distillated. The residue is analyzed by carrier distillation and by total burning techniques. The analytical results are discussed and compared with those obtained destroying the organic matter without carrier and using the copper spark technique. (Author) 12 refs.

  18. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

  19. Multiregion reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

  20. The Role of Non-Destructive Testing in Test-Reactor Operation at the National Reactor Testing Station; Role des Essais Non Destructifs dans l'Exploitation des Reacteurs d'Essai au Centre National d'Essais de Reacteurs; Rol' nedestruktivnykh ispytanij pri ehkspluatatsii ispytatel'nykh reaktorov na natsional'noj stantsii po ispytaniyam reaktorov; Papel de los Metodos No Destructivos en la Explotacion de los Reactores de la National Reactor Testing Station

    Energy Technology Data Exchange (ETDEWEB)

    Francis, W. C.; Brown, E. S.; Burdick, E. E.; Gibson, G. W.; Tingey, F. H. [Phillips Petroleum Company, Atomic Energy Division, Idaho Falls, Idaho (United States)

    1965-10-15

    'un densimetre, permettent de determiner la distribution du combustible. On a habituellement recours a la radiographie des soudures pour les parties constitutives des reacteurs et des boucles d'essai. Le dispositif perfectionne de mesure de la reactivite (Advanced Reactivity Measurement Facility, ARMF) permet de determiner, pour chaque cycle de reacteur, l'irradiation du combustible et l'empoisonnement dans des specimens. Une application assez peu courante pour un assemblage critique est la mesure de la teneur en bore du combustible dans l'assemblage critique d'essai en genie des reacteurs (Engineering Test Reactor Critical Facility, ETRC). Le controle par courants de Foucault et par des procedes mecaniques de l'espacement des plaques de combustible et la mesure par courants de Foucault de l'epaisseur de l'oxydation (corrosion) sur les plaques irradiees ont donne d'excellents resultats. Des methodes complementaires qui ont fait leurs preuves sont l'inspection par liquide penetrant et les essais a l'azote liquide pour les craquelures superficielles, les essais par recuit thermique pour les souitlures et l'exploration par rayons gamma des plaques irradiees. On a recours a l'essai hydraulique d'un echantillon statistique d'elements combustibles pour verifier l'integrite structurale, notamment la resistance de la liaison entre les plaques de combustible et la gaine. Des efforts constants sont deployes pour ameliorer les methodes actuelles et mettre au point de nouveaux procedes de controle non destructif. (author) [Spanish] Los reactores de ensayo de la National Reactor Testing Station suponen una enorme inversion (superior a 100 millones de dolares) y la necesidad de explotarlos en condiciones de seguridad obliga a proceder a un control de calidad muy estricto de los componentes nucleares y de ensayo, especialmente en lo que respecta a los elementos combustibles y de control. Por tanto, los metodos no. destructivos son fundamentales para determinar la calidad de estos componentes

  1. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    In the system described the fuel elements are arranged vertically in groups and are supported in such a manner as to tend to tilt them towards the center of the respective group, the fuel elements being urged laterally into abutment with one another. The elements have interlocking bearing pads, whereby lateral movement of adjacent elements is resisted; this improves the stability of the reactor core during refuelling operations. Fuel elements may comprise clusters of parallel fuel pins enclosed in a wrapper of hexagonal cross section, with bearing pads in the form of spline-like ribs located on each side of the wrapper and extending parallel to the longitudinal axis of the fuel element, being interlockable with ribs on pads of adjacent fuel elements. The arrangement is applicable to a reactor core in which fuel elements and control rod guide tubes are arranged in modules each of which comprises a cluster of at least three fuel elements, one of which is rigidly supported whilst the others are resiliently tilted towards the center of the cluster so as to lean on the rigidly supported element. It is also applicable to modules comprising a cluster of six fuel elements, each resiliently tilted towards a central void to form a circular arch. The modules may include additional fuel elements located outside the clusters and also resiliently tilted towards the central voids, the latter being used to accommodate control rod guide tubes. The need for separate structural members to act as leaning posts is thus avoided. Such structural members are liable to irradiation embrittlement, that could lead to core failure. (U.K.)

  2. REMOVAL OF ORGANIC MATTER AND TOXICITY IN AN UPFLOW IMMOBILIZED BIOMASS ANAEROBIC REACTOR TREATING HOSPITAL WASTEWATER: PRELIMINARY EVALUATION

    Directory of Open Access Journals (Sweden)

    MÓNICA PORRAS TORRES

    2013-01-01

    Full Text Available El objetivo de esta investigación consistió en evaluar el desempeño de un reactor anaerobio de flujo ascendente de biomasa inmovilizada (RAFABI tratando un efluente hospitalario real. Se estudió la remoción de materia orgánica y toxicidad, por medio de análisis como UV254, DQOfiltrada y determinación del porcentaje de inhibición en el crecimiento de la raíz de la cebolla. Los resultados mostraron que el proceso biológico estuvo estable durante los 287 días de operación continua, el valor medio de la relación AI/AP fue de 1.21±0.08, indicando que no hubo acumulación de ácidos en el sistema. Sin embargo, los valores de la eficiencia de remoción de DQOfiltrada, 56±15% y UV254, 21±36%, no fueron representativos. La toxicidad se redujo en 50%. Con base en lo anterior, es necesario utilizar el reactor anaerobio en combinación con otros procesos como por ejemplo los procesos de oxidación avanzada, para continuar reduciendo la materia orgánica recalcitrante al proceso anaerobio. Se comprobó la capacidad que tienen los reactores anaerobios de biomasa inmovilizada para remover la toxicidad.

  3. Reactor building

    International Nuclear Information System (INIS)

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

    In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

  4. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

    A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

  5. Reactor physics challenges in GEN-IV reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, Michael K.; Hejzlar, Pavel [Massachusetts Institute of Technology, MA (United States)

    2005-02-15

    An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources.

  6. Reactor physics challenges in GEN-IV reactor design

    International Nuclear Information System (INIS)

    Driscoll, Michael K.; Hejzlar, Pavel

    2005-01-01

    An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources

  7. Nuclear reactor plant with a gas-cooled nuclear reactor situated in a cylindrical prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Becker, G.; Elter, C.; Fritz, R.; Rautenberg, J.; Schoening, J.; Stracke, W.

    1986-01-01

    A simplified construction of the nuclear reactor plant with a guarantee of great safety is achieved by the auxiliary heat exhangers, which remove the post-shutdown heat in fault situations, being arranged in the wellknown way in pairs above one another in a vertical shaft. The associated auxiliary blowers are situated at the top for the upper auxiliary heat exchangers and at the bottom for the lower auxiliary heat exchangers. The cold gas is taken from the lower auxiliary blowers through a parallel gas pipe laid in concrete, which enters the vertical shaft concerned in the area of the cold gas pipe. (orig./HP) [de

  8. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  9. Development of the user interface for visualization of the auxiliary systems of the TRIGA Mark III nuclear reactor; Desarrollo de la interface de usuario para la visualizacion de los sistemas auxiliares del reactor nuclear Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Merced D, J. E.

    2016-07-01

    The Instituto Nacional de Investigaciones Nucleares (ININ) has a nuclear research reactor type swimming pool with movable core cooled and moderate with light water. The nominal maximum power of the reactor is 1 MW in steady-state operation and can be pulsed at a maximum power of 2,000 MW for approximately 10 milliseconds. This reactor is mainly used to study the effects of radiation on various materials and substances. In 2001 the new control console of the nuclear reactor was installed which was based on two digital computers, one computer controls the bar management mechanisms and the other the systems to the reactor operator. In 2004, the control computer was replaced and the software was updated. Within the modernization and/or updating of the TRIGA Mark III reactor of ININ, is intended (theme of this work) to develop the user interface for the visualization of the auxiliary systems, through a Man-Machine Interface module for the renewal process of the control console. The man-machine interface system to be developed will have communication with the programmable logic controllers that will be constantly monitored and controlled to obtain real-time variables of the reactor behavior. (Author)

  10. Análisis teórico de la influencia del régimen de mezclado para la producción de acetato de etilo

    Directory of Open Access Journals (Sweden)

    Hugo Alexander Martínez C.

    1999-01-01

    Full Text Available Se efectúa un análisis teórico a la reacción de esteriticación entre ácido acético y etanol tendiente a conocer la conveniencia de operar en modo de CSTR (reactor de mezcla total ideal o PFTR (reactor de flujo pistón, para un conjunto dado de parámetros fisicoquímicos y cinética de esterificación de seudoprlmer orden.

  11. The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1983-01-01

    A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

  12. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

    Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

  13. Cuando todo es político, ¿qué es la política?: una acotación empírica desde el post-humanismo Cuando todo es político, ¿qué es la política?: una acotación empírica desde el post-humanismo

    Directory of Open Access Journals (Sweden)

    Fernando Calonge Reíllo

    2009-04-01

    Full Text Available Este artículo intenta contribuir a resolver una de las más marcadas deficiencias que se atribuyen al paradigma post-humanista. En la medida en que desde dicho paradigma se señala que la ontología es política, se dificulta al mismo tiempo la comprensión de un sentido particular y específico de política. Desde esta visión, todo es político y, en consecuencia, nada es político. En el artículo reviso algunas de las soluciones que se han aportado para solventar esta carencia y añado las limitaciones que presentan. Buscando una vía para evitar la acusación, ofrezco una lectura post-humanista de las migraciones, para el caso concreto de las mujeres inmigrantes en la Comunidad de Madrid.Apartir de esa lectura, planteo la interrogación sobre qué constituyó, en concreto, la especificidad de las mujeres que migraron por razones políticas. A raíz de esta elucidación induzco unos mínimos criterios para caracterizar a la ‘migración política’, que la diferencie de otros tipos de migraciones. Desde esta elucidación realizo una extrapolación al campo más general de la política para el post-humanismo. Habiendo elucidado el hecho político de las migraciones políticas, intento establecer, dentro del paradigma post-humanista, qué constituye finalmente una noción específica y concreta sobre lo político que subsane la crítica de partida sobre la falta de especificidad de la política para el post-humanismo.In this article I will try to solve one of the most acute deficiencies in the post-humanist paradigm. As it is known, in this paradigm it is argued that ontology is political, therefore it has to be concluded that everything is political. In that case, post-humanism is accused of lacking a specific sense for the term ‘politics’. In the article, I review some of the solutions that have been proposed to give a more specific sense of politics and I include the reasons why such solutions are inadequate. Searching for a way to

  14. Utilization of vinasse effluents from an anaerobic reactor

    Energy Technology Data Exchange (ETDEWEB)

    Costa, F J.C.B.; Rocha, B B.M.; Viana, C E; Toledo, A C

    1986-01-01

    An anaerobic reactor was developed to biodigest alcohol distillery wastes. A further post-treatment of the effluent reduced the level of pollution to the point of eventually discharging into streams and rivers. The present work also analyses the use of biodigested vinasse as a source of food for fish. Very high efficiencies were obtained during primary and secondary treatment of vinasse effluent, as demonstrated by the greatly reduced organic load. The utilization of the treated effluent as a source of fish food presents an excellent alternative for the Brazilian alcohol industry. (Refs. 6).

  15. Kits para aerossol em um serviço de saúde: uma análise microbiológica após reprocessamento Kits para aerosol en un servicio de salud: un análisis microbiológico post procesamiento Aerosol kits in a health service: a post-processing microbiologic analysis

    Directory of Open Access Journals (Sweden)

    Patrícia Staciarini Anders

    2008-06-01

    Full Text Available Kits para aerossol são artigos utilizados na terapêutica de afecções do trato respiratório e requerem no mínimo desinfecção de nível intermediário para reuso. Os objetivos deste estudo foram verificar uma possível contaminação microbiana em kits para aerossol pós-reprocessamento e identificar os microrganismos isolados. Estudo transversal, experimental realizado na unidade pediátrica de um hospital em Goiânia-GO. Coletaram-se amostras de três segmentos (máscara, copo, interior da extensão de 15 kits previamente desinfetados, que foram semeadas em diferentes meios de culturas e os microrganismos isolados foram identificados por provas bioquímicas. Dos 15 kits analisados, 13 copos, nove extensões e 13 máscaras estavam contaminados. Isolou-se no total 101 UFC, 39 provindos dos copos, 20 das extensões e 42 das máscaras. Dentre os patógenos isolados destaca-se: Staphylococcus coagulase positivo, Staphylococcus coagulase negativo, Bastonetes Gram negativo fermentadores, Bastonetes Gram negativo não fermentadores, micrococose leveduras. A detecção microbiana indica prováveis falhas no reprocessamento desses artigos.Kits para aerosol son artículos utilizados en la terapéutica de afecciones del tracto respiratorio y requieren en lo mínimo desinfección a nivel intermediario para volver a ser usado. Los objetivos de este estudio fueron verificar una posible contaminación microbiológica en kits para aerosol post-reprocesamiento e identificar los microorganismos aislados. Se trata de un estudio transversal, experimental realizado en la unidad pediátrica de un hospital en Goiânia-GO. Se recolectaron muestras de tres segmentos (máscara, vaso, interior de la extensión de 15 kits previamente desinfectados, que fueron sembradas en diferentes medios de cultivo y los microorganismos aislados fueron identificados por pruebas bioquímicas. De los 15 kits analizados, 13 vasos, nueve extensiones y 13 máscaras estaban contaminados

  16. Multi-physics design and analyses of long life reactors for lunar outposts

    Science.gov (United States)

    Schriener, Timothy M.

    Future human exploration of the solar system is likely to include establishing permanent outposts on the surface of the Moon. These outposts will require reliable sources of electrical power in the range of 10's to 100's of kWe to support exploration and resource utilization activities. This need is best met using nuclear reactor power systems which can operate steadily throughout the long ˜27.3 day lunar rotational period, irrespective of location. Nuclear power systems can potentially open up the entire lunar surface for future exploration and development. Desirable features of nuclear power systems for the lunar surface include passive operation, the avoidance of single point failures in reactor cooling and the integrated power system, moderate operating temperatures to enable the use of conventional materials with proven irradiation experience, utilization of the lunar regolith for radiation shielding and as a supplemental neutron reflector, and safe post-operation decay heat removal and storage for potential retrieval. In addition, it is desirable for the reactor to have a long operational life. Only a limited number of space nuclear reactor concepts have previously been developed for the lunar environment, and these designs possess only a few of these desirable design and operation features. The objective of this research is therefore to perform design and analyses of long operational life lunar reactors and power systems which incorporate the desirable features listed above. A long reactor operational life could be achieved either by increasing the amount of highly enriched uranium (HEU) fuel in the core or by improving the neutron economy in the reactor through reducing neutron leakage and parasitic absorption. The amount of fuel in surface power reactors is constrained by the launch safety requirements. These include ensuring that the bare reactor core remains safely subcritical when submerged in water or wet sand and flooded with seawater in the unlikely

  17. Diseño y simulación de un reactor prototipo que soporte condiciones de hidrogenación para crudo pesado con una capacidad de 1 galón por segundo para Petroecuador

    OpenAIRE

    Checa Ramírez, Pablo Andrés; Andy Cerda, Amilkar Patricio

    2010-01-01

    El presente proyecto se plantea en cinco capítulos estrictamente relacionados: Procesos teóricos, Tipos de reactores, Diseño y cálculo del reactor prototipo, Método del cálculo de prueba y error, y, costos que genera el diseño

  18. Arranque y operación a escala real de un sistema de tratamiento de lodos activos para aguas residuales de matadero

    OpenAIRE

    Pabón, Sandra Liliana; Suárez Gélvez, John Hermógenes

    2010-01-01

    La investigación se desarrolló en la planta de tratamiento de aguas residuales de la empresa Frigorífico Frigofrontera Ltda., la cual está constituida por un sistema de tratamiento primario que incluye cribado, desarenado, trampa de grasas y sedimenta- ción, un reactor de lodo activo con recirculación para el tratamiento secundario y un filtro descendente como tratamiento ter- ciario. El caudal tratado en promedio es de 1,38 L/s; el reactor de lodo activo tiene un volumen de 144 m3, un ...

  19. Sodium fast reactor power monitoring using gamma spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R.; Normand, S.; Barbot, L.; Domenech, T.; Kondrasovs, V.; Corre, G.; Frelin, A.M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, CEA - Saclay DRT/LIST/DETECS/SSTM, Batiment 516 - P.C. no 72, Gif sur Yvette, F-91191 (France); Montagu, T.; Dautremer, T.; Barat, E. [CEA, LIST, Laboratoire Processus Stochastiques et Spectres (France); Ban, G. [ENSICAEN (France)

    2009-06-15

    This work deals with the use of high flux gamma spectrometry to monitor the fourth generation of sodium fast reactor (SFR) power. The simulation study part of this work has shown that power monitoring in a short time response and with a good accuracy is possible. An experimental test is under preparation at the French SFR Phenix experimental reactor to validate simulation studies. First, physical calculations have been done to correlate gamma activity to the released thermal power. Gamma emitter production rate in the reactor core was calculated with technical and nuclear data as the sodium velocity, the atomic densities, Phenix neutron spectrum and incident neutron cross-sections of reactions producing gamma emitters. A thermal hydraulic transfer function was used for modeling primary sodium flow in our calculations. For the power monitoring problematic, use of a short decay period gamma emitter will allow to have a very fast response system without cumulative effect. We have determined that the best tagging agent is 20F which emits 1634 keV energy photons with a decay period of 11 s. The gamma spectrum was determined by flux point and a pulse high tally MCNP5.1.40 simulation and shown the possibility to measure the signal of this radionuclide. The experiment will be set during the reactor 'end life testing'. The Delayed Neutron Detection (DND) room has been chosen as the best available location on Phenix reactor to measure this kind of radionuclide due to a short transit time from reactor core to measurement sample. This location is optimum for global power measurement because homogenized sampling in the reactor hot pool. The main spectrometer is composed of a coaxial high purity germanium diode (HPGe) coupled with a transistor reset preamplifier. The HPGe diode signal will be processed by the Adonis digital signal processing due to high flux and fast activity measurement. Post-processing softwares will be used to limit statistical problems of the

  20. Calculation of reactivity for safety in nuclear reactors; Calculo de la reactividad para seguridad en reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Suescun D, D. [Universidad Surcolombiana, Av. Pastrana Borrero - Carrera 1, Neiva, Huila (Colombia); Rojas A, O., E-mail: daniel.suescun@usco.edu.co [Universidad Popular Autonoma del Estado de Puebla, Av. 9 Pte 1908, Barrio de Santiago, 72410 Puebla (Mexico)

    2017-09-15

    The measurement of reactivity is a function of time and its calculation results from the variation in nuclear power from the inverse equation of punctual kinetics. This equation is a differential integral, where the term of the integral conserves the historical power and the differential part is directly related to the period of the reactor. In practice, in a nuclear plant, sensors are required to record the signals. For example, the movements of the control rods that cause the fluctuations of nuclear power over time commonly generate signals with noise, an event that makes difficult to estimate the reactivity. Thus is necessary and very useful to build digital reactivity meters in real time, since allows a reactor to be operated with greater security. The calculation of the reactivity is carried out using punctual kinetics, especially the concentration of delayed neutron precursors. In this work we present a new way to reduce the fluctuations in the calculation of the reactivity, for the high precision we propose the generalization of the predictor and corrector of the Adams-Bashforth-Moulton (ABM) method of order 4 to solve numerically the equations of the point kinetics for the calculation of the reactivity, without using the power history, due to the nature of the equations of the punctual kinetics, the modifiers of the different predictors are used to increase the accuracy in the approximation obtained accompanied by the filter known as Savitzky-Golay (Sg), allow to reduce the fluctuations of reactivity. It is known that the Sg filter softens and does not attenuate the nuclear power regardless of its shape, guarantees to reduce noise levels up to σ = 0.01, with a calculation time step of σ = 0.01, s. This formulation uses a polynomial approximation of Gram, with a degree d = 2, to calculate the convolution coefficients by means of an analytical formula that is implemented computationally and avoids problems of bad conditioning, caused by the inversion of a